Control-oriented Automatic System for Transport Analysis (ASTRA)-Matlab integration for Tokamaks

International Nuclear Information System (INIS)

Sevillano, M.G.; Garrido, I.; Garrido, A.J.

2011-01-01

The exponential growth in energy consumption has led to a renewed interest in the development of alternatives to fossil fuels. Between the unconventional resources that may help to meet this energy demand, nuclear fusion has arisen as a promising source, which has given way to an unprecedented interest in solving the different control problems existing in nuclear fusion reactors such as Tokamaks. The aim of this manuscript is to show how one of the most popular codes used to simulate the performance of Tokamaks, the Automatic System For Transport Analysis (ASTRA) code, can be integrated into the Matlab-Simulink tool in order to make easier and more comfortable the development of suitable controllers for Tokamaks. As a demonstrative case study to show the feasibility and the goodness of the proposed ASTRA-Matlab integration, a modified anti-windup Proportional Integral Derivative (PID)-based controller for the loop voltage of a Tokamak has been implemented. The integration achieved represents an original and innovative work in the Tokamak control area and it provides new possibilities for the development and application of advanced control schemes to the standardized and widely extended ASTRA transport code for Tokamaks. -- Highlights: → The paper presents a useful tool for rapid prototyping of different solutions to deal with the control problems arising in Tokamaks. → The proposed tool embeds the standardized Automatic System For Transport Analysis (ASTRA) code for Tokamaks within the well-known Matlab-Simulink software. → This allows testing and combining diverse control schemes in a unified way considering the ASTRA as the plant of the system. → A demonstrative Proportional Integral Derivative (PID)-based case study is provided to show the feasibility and capabilities of the proposed integration.

ASTRA - an automatic system for transport analysis in a tokamak

International Nuclear Information System (INIS)

Pereverzev, G.V.; Yushmanov, P.N.; Dnestrovskii, A.Yu.; Polevoi, A.R.; Tarasjan, K.N.; Zakharov, L.E.

1991-08-01

The set of codes described here - ASTRA (Automatic System of Transport Analysis) - is a flexible and effective tool for the study of transport mechanisms in reactor-oriented facilities of the tokamak type. Flexibility is provided within the ASTRA system by a wide choice of standard relationships, functions and subroutines representing various transport coefficients, methods of auxiliary heating and other physical processes in the tokamak plasma, as well as by the possibility of pre-setting transport equations and variables for data output in a simple and conseptually transparent form. The transport code produced by the ASTRA system provides an adequate representation of the discharges for present experimental conditions. (orig.)

System studies for quasi-steady-state advanced physics tokamak

International Nuclear Information System (INIS)

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

1983-11-01

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

Development of Tokamak reactor system code and conceptual studies of DEMO with He Cooled Molten Li blanket

International Nuclear Information System (INIS)

Hong, B.G.; Lee, Dong Won; Kim, Yong Hi

2007-01-01

To develop the concepts of fusion power plants and identify the design parameters, we have been developing the tokamak reactor system code. The system code can take into account a wide range of plasma physics and technology effects simultaneously and it can be used to find design parameters which optimize the given figure of merits. The outcome of the system studies using the system code is to identify which areas of plasma physics and technologies and to what extent should be developed for realization of a given fusion power plant concepts. As an application of the tokamak reactor system code, we investigate the performance of DEMO for early realization with a limited extension from the plasma physics and technology used in the design of the ITER. Main requirements for DEMO are selected as: 1) to demonstrate tritium self-sufficiency, 2) to generate net electricity, and 3) for steady-state operation. The size of plasma is assumed to be same as that of ITER and the plasma parameters which characterize the performance, i.e. normalized β value, β N , confinement improvement factor for the H-mode, H and the ratio of the Greenwald density limit n/n G are assumed to be improved beyond those of ITER: β N >2.0, H>1.0 and n/n G >1.0. Tritium self-sufficiency is provided by the He Cooled Molten Lithium (HCML) blanket with the total thickness of 2.5 m including the shield. With n/n G >1.2, net electric power bigger than 500 MW is possible with β N >4.0 andH>1.2. To access operation space for higher electric power, main restrictions are given by the divertor heat load and the steady-state operation requirements. Developments in both plasma physics and technology are required to handle high heat load and to increase the current drive efficiency. (orig.)

The GBS code for tokamak scrape-off layer simulations

International Nuclear Information System (INIS)

Halpern, F.D.; Ricci, P.; Jolliet, S.; Loizu, J.; Morales, J.; Mosetto, A.; Musil, F.; Riva, F.; Tran, T.M.; Wersal, C.

2016-01-01

We describe a new version of GBS, a 3D global, flux-driven plasma turbulence code to simulate the turbulent dynamics in the tokamak scrape-off layer (SOL), superseding the code presented by Ricci et al. (2012) [14]. The present work is driven by the objective of studying SOL turbulent dynamics in medium size tokamaks and beyond with a high-fidelity physics model. We emphasize an intertwining framework of improved physics models and the computational improvements that allow them. The model extensions include neutral atom physics, finite ion temperature, the addition of a closed field line region, and a non-Boussinesq treatment of the polarization drift. GBS has been completely refactored with the introduction of a 3-D Cartesian communicator and a scalable parallel multigrid solver. We report dramatically enhanced parallel scalability, with the possibility of treating electromagnetic fluctuations very efficiently. The method of manufactured solutions as a verification process has been carried out for this new code version, demonstrating the correct implementation of the physical model.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Computer code for the costing and sizing of TNS tokamaks

International Nuclear Information System (INIS)

Sink, D.A.; Iwinski, E.M.

1977-01-01

A FORTRAN code for the COsting And Sizing of Tokamaks (COAST) is described. The code was written to conduct detailed analyses on the engineering features of the next tokamak fusion device following TFTR. The ORNL/Westinghouse study of TNS (The Next Step) has involved the investigation of a number of device options, each over a wide range of plasma sizes. A generalized description of TNS is incorporated in the code and includes refined modeling of over forty systems and subsystems. Considerable detailed design and analyses have provided the basis for the thermal, electrical, mechanical, nuclear, chemical, vacuum, and facility engineering of the various subsystems. Currently, the code provides a tool for the systematic comparison of four toroidal field (TF) coil technologies allowing both D-shaped and circular coils. The coil technologies are: (1) copper (both room temperature and liquid-nitrogen cooled), (2) superconducting NbTi, (3) superconducting Nb 3 Sn, and (4) a Cu/NbTi/ hybrid. For the poloidal field (PF) coil systems copper conductors are assumed. The ohmic heating (OH) coils are located within the machine bore and have an air core, while the shaping field (SF) coils are located either within or outside the TF coils. The PF coil self and mutual inductances are calculated from the geometry, and the PF coil power supplies are modeled to account for time-dependent profiles for voltages and currents as governed by input data. Plasma heating is assumed to be by neutral beams, and impurity control is either passive or by a poloidal divertor system. The size modeling allows considerable freedom in specifying physics assumptions, operating scenarios, TF operating margin, and component geometric and performance parameters. Cost relationships have been developed for both plant and capital equipment and for annual utility and fuel expenses. The code has been used successfully to reproduce the sizing and costing of TFTR in order to calibrate the various models

System studies of compact ignition tokamaks

International Nuclear Information System (INIS)

Galambos, J.D.; Peng, Y.K.M.; Blackfield, D.T.

1986-01-01

A new version of the FEDC Tokamak System Code (TSC) has been developed to analyze the Compact Ignition Tokamak (CIT). These proposed experiments have small (major radius F 1.5m) and high magnetic fields (B J 10T), and are characterized by reduced cost. Key design constraints of CIT include limits to the high stress levels in the magnetic coils, limits to the large temperature rises in the coils and on the first wall or divertor plate, minimizing power supply requirements, and assuring adequate plasma performance in fusion ignition and burn time consistent with the latest physics understanding. We present systems code level studies of CIT parameter space here for a range of design options with various design constraints. The present version of the TSC incorporates new models for key components of CIT. For example, new algorithms have been incorporated for calculating stress levels in the TFC and ohmic solenoid, temperature rise in the magnetic coils, peak power requirements, plasma MHD equilibrium and volt-second capability. The code also incorporates a numerical optimizer to find combinations of engineering quantities (device size, coil sizes, coil current densities etc.) and physics quantities (plasma density temperature, and beta, etc.) which satisfy all the constraints and can minimize or maximize a figure of merit (e.g., the major radius). This method was recently used in a mirror reactor system code (3) for the Minimara concept development

pTSC: Data file editing for the Tokamak Simulation Code

International Nuclear Information System (INIS)

Meiss, J.D.

1987-09-01

The code pTSC is an editor for the data files needed to run the Princeton Tokamak Simulation Code (TSC). pTSC utilizes the Macintosh interface to create a graphical environment for entering the data. As most of the data to run TSC consists of conductor positions, the graphical interface is especially appropriate

Study of Globus-M Tokamak Poloidal System and Plasma Position Control

Science.gov (United States)

Dokuka, V. N.; Korenev, P. S.; Mitrishkin, Yu. V.; Pavlova, E. A.; Patrov, M. I.; Khayrutdinov, R. R.

2017-12-01

In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.

Whist code calculations of ignition margin in an ignition tokamak

International Nuclear Information System (INIS)

Sheffield, J.

1985-01-01

A simple global model was developed to determine the ignition margin of tokamaks including electron and ion conduction losses. A comparison of this model with results from a 1 1/2 dimensional Whist code is made

TSC [Tokamak Simulation Code] disruption scenarios and CIT [Compact Ignition Tokamak] vacuum vessel force evolution

International Nuclear Information System (INIS)

Sayer, R.O.; Peng, Y.K.M.; Strickler, D.J.; Jardin, S.C.

1990-01-01

The Tokamak Simulation Code and the TWIR postprocessor code have been used to develop credible plasma disruption scenarios for the Compact Ignition Tokamak (CIT) in order to predict the evolution of forces on CIT conducting structures and to provide results required for detailed structural design analysis. The extreme values of net radial and vertical vacuum vessel (VV) forces were found to be F R =-12.0 MN/rad and F Z =-3.0 MN/rad, respectively, for the CIT 2.1-m, 11-MA design. Net VV force evolution was found to be altered significantly by two mechanisms not noted previously. The first, due to poloidal VV currents arising from increased plasma paramagnetism during thermal quench, reduces the magnitude of the extreme F R by 15-50% and modifies the distribution of forces substantially. The second effect is that slower plasma current decay rates give more severe net vertical VV loads because the current decay occurs when the plasma has moved farther from midplane than is the case for faster decay rates. 7 refs., 9 figs., 1 tab

Tokamak Simulation Code modeling of NSTX

International Nuclear Information System (INIS)

Jardin, S.C.; Kaye, S.; Menard, J.; Kessel, C.; Glasser, A.H.

2000-01-01

The Tokamak Simulation Code [TSC] is widely used for the design of new axisymmetric toroidal experiments. In particular, TSC was used extensively in the design of the National Spherical Torus eXperiment [NSTX]. The authors have now benchmarked TSC with initial NSTX results and find excellent agreement for plasma and vessel currents and magnetic flux loops when the experimental coil currents are used in the simulations. TSC has also been coupled with a ballooning stability code and with DCON to provide stability predictions for NSTX operation. TSC has also been used to model initial CHI experiments where a large poloidal voltage is applied to the NSTX vacuum vessel, causing a force-free current to appear in the plasma. This is a phenomenon that is similar to the plasma halo current that sometimes develops during a plasma disruption

Comparison benchmark between tokamak simulation code and TokSys for Chinese Fusion Engineering Test Reactor vertical displacement control design

International Nuclear Information System (INIS)

Qiu Qing-Lai; Xiao Bing-Jia; Guo Yong; Liu Lei; Wang Yue-Hang

2017-01-01

Vertical displacement event (VDE) is a big challenge to the existing tokamak equipment and that being designed. As a Chinese next-step tokamak, the Chinese Fusion Engineering Test Reactor (CFETR) has to pay attention to the VDE study with full-fledged numerical codes during its conceptual design. The tokamak simulation code (TSC) is a free boundary time-dependent axisymmetric tokamak simulation code developed in PPPL, which advances the MHD equations describing the evolution of the plasma in a rectangular domain. The electromagnetic interactions between the surrounding conductor circuits and the plasma are solved self-consistently. The TokSys code is a generic modeling and simulation environment developed in GA. Its RZIP model treats the plasma as a fixed spatial distribution of currents which couple with the surrounding conductors through circuit equations. Both codes have been individually used for the VDE study on many tokamak devices, such as JT-60U, EAST, NSTX, DIII-D, and ITER. Considering the model differences, benchmark work is needed to answer whether they reproduce each other’s results correctly. In this paper, the TSC and TokSys codes are used for analyzing the CFETR vertical instability passive and active controls design simultaneously. It is shown that with the same inputs, the results from these two codes conform with each other. (paper)

The modeling of the RF system performance in TCA/BR tokamak

International Nuclear Information System (INIS)

Ruchko, L.; Galvao, R.M.O.; Nascimento, I.; Ozono, E.; Lerche, E.; Degasperi, F.T.; Tuszel, A.G.

1996-01-01

The results of numerical simulation of RF Alfven wave heating system that is intended to be used in TCA/BR tokamak are presented. The problem of monochromatic travelling RF field excitation in TCA/BR tokamak is analyzed by means of numerical simulation. The spectrum of the excited Alfven waves is determined using a one-dimensional MHD code. The transient time and AC analysis of the RF generator performance with antenna loading are discussed. (author). 9 refs., 6 figs

KDAS: General-Purpose Data Acquisition System Developed for KAIST-Tokamak

International Nuclear Information System (INIS)

Seo, Seong-Heon; Choe, Wonho; Chang, Hong-Young; Jeong, Seung-Ho

2000-01-01

The Korea Advanced Institute of Science and Technology (KAIST)-Tokamak Data Acquisition System (KDAS) was originally developed for KAIST-Tokamak (R/a = 0.53 m/0.14 m). It operates on a distributed system based on personal computers and has a driver-based hierarchical structure. Since KDAS can be dynamically composed of any number of available computers, and the hardware-dependent codes can be thoroughly separated into external drivers, it exhibits excellent system performance flexibility and extensibility and can optimize various user needs. It collectively controls the VXI, CAMAC, GPIB, and RS232 instrument hybrids. With these useful and convenient features, it can be applied to any computerized experiment, especially to fusion-related research. The system design and features are discussed in detail

Tokamak plasma power balance calculation code (TPC code) outline and operation manual

International Nuclear Information System (INIS)

Fujieda, Hirobumi; Murakami, Yoshiki; Sugihara, Masayoshi.

1992-11-01

This report is a detailed description on the TPC code, that calculates the power balance of a tokamak plasma according to the ITER guidelines. The TPC code works on a personal computer (Macintosh or J-3100/ IBM-PC). Using input data such as the plasma shape, toroidal magnetic field, plasma current, electron temperature, electron density, impurities and heating power, TPC code can determine the operation point of the fusion reactor (Ion temperature is assumed to be equal to the electron temperature). Supplied flux (Volt · sec) and burn time are also estimated by coil design parameters. Calculated energy confinement time is compared with various L-mode scaling laws and the confinement enhancement factor (H-factor) is evaluated. Divertor heat load is predicted by using simple scaling models (constant-χ, Bohm-type-χ and JT-60U empirical scaling models). Frequently used data can be stored in a 'device file' and used as the default values. TPC code can generate 2-D mesh data and the POPCON plot is drawn by a contour line plotting program (CONPLT). The operation manual about CONPLT code is also described. (author)

Engineering parameters for four ignition TNS tokamak reactor systems

International Nuclear Information System (INIS)

Varljen, T.C.; Gibson, G.; French, J.W.; Heck, F.M.

1977-01-01

The ORNL/Westinghouse program for The Next Step (TNS) tokamak beyond TFTR has examined a large number of potential configurations for D-T burning ignition tokamak systems. An objective of this work has been to quantify the trade-offs associated with the assumption of certain plasma physics criteria and toroidal field coil technologies. Four tokamak system point designs are described, each representative of the TF coil technologies considered, to illustrate the engineering features associated with each concept. Point designs, such as the ones discussed herein, have been used to develop component size, performance and cost scaling relationships which have been incorporated in a digital computer code to facilitate an examination of the total design and cost impact of candidate design approaches. The point designs which are described are typical, however, they have not been individually optimized. The options are distinguished by the TF coil technology chosen and include: (1) a high field water-cooled copper TF system, (2) a moderate field NbTi superconducting TF system, (3) a high field Nb 3 Sn superconducting TF system, and (4) a high field hybrid TF system with outer NbTi superconducting windings and inner water-cooled copper windings. Descriptions are provided for the major device components and all major support systems including power supplies, vacuum systems, fuel systems, heat transport and facility systems

Numerical study for determining PF coil system parameters in MHD equilibrium of KT-2 tokamak plasma

International Nuclear Information System (INIS)

Ryu, J.; Hong, S.H.; Lee, K.W.; Hong, B.G.; In, S.R.; Kim, S.K.

1995-01-01

The KT-2 is a large-aspect-ratio medium-sized divertor tokamak in the conceptual design phase and planned to be operational in 1998 at the Korea Atomic Energy Research Institute (KAERI). Plasma equilibrium in tokamak can be acquired by controlling the current of poloidal field (PF) coils in appropriate geometries and positions. In this study, the authors have performed numerical calculations to achieve the various equilibrium conditions fitting given plasma shapes and satisfying PF current limitations. Usually an ideal magnetohydrodynamic (MHD) equation is used to obtain the equilibrium solution of tokamak plasma, and it is practical to take advantage of a numerical method in solving the MHD equation because it has nonlinear source terms. Two equilibrium codes have been applied to find a double-null configuration of free-boundary tokamak plasma in KT-2: one is of the authors' own developing and the other is a free-boundary tokamak equilibrium code (FBT) that has been used mainly for the verification of developed code's results. PF coil system parameters including their positions and currents are determined for the optimization of input power required when the specifications of KT-2 tokamak are met. Then, several sets of equilibrium conditions during the tokamak operation are found to observe the changes of poloidal field currents with the passing of operation time step, and the basic stability problems related with the magnetic field structure is also considered

Improved core-edge tokamak transport simulations with the CORSICA 2 code

International Nuclear Information System (INIS)

Tarditi, A.; Cohen, R.H.; Crotinger, J.A.

1996-01-01

The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented

Plasmator. A numerical code for simulation of plasma transport in Tokamaks

International Nuclear Information System (INIS)

Guasp, J.

1979-01-01

Plasmator is a flexible monodimensional numerical code for plasma transport in Tokamaks of circular cross-section, it allows neutral particle transport and impurity effects. The code leaves a total freedom in the analytical form of transport coefficients. It has been writen in Fortran-V for the UNIVAC-1100/80 from JEN and allows for the possibility of graphics for radial profiles and temporal evolution of the main plasma magnitudes, as well in three-dimensional as in two-dimensional representation either on a Calcomp plotter or in the printer. (author)

CXSFIT Code Application to Process Charge-Exchange Recombination Spectroscopy Data at the T-10 Tokamak

Science.gov (United States)

Serov, S. V.; Tugarinov, S. N.; Klyuchnikov, L. A.; Krupin, V. A.; von Hellermann, M.

2017-12-01

The applicability of the CXSFIT code to process experimental data from Charge-eXchange Recombination Spectroscopy (CXRS) diagnostics at the T-10 tokamak is studied with a view to its further use for processing experimental data at the ITER facility. The design and operating principle of the CXRS diagnostics are described. The main methods for processing the CXRS spectra of the 5291-Å line of C5+ ions at the T-10 tokamak (with and without subtraction of parasitic emission from the edge plasma) are analyzed. The method of averaging the CXRS spectra over several shots, which is used at the T-10 tokamak to increase the signal-to-noise ratio, is described. The approximation of the spectrum by a set of Gaussian components is used to identify the active CXRS line in the measured spectrum. Using the CXSFIT code, the ion temperature in ohmic discharges and discharges with auxiliary electron cyclotron resonance heating (ECRH) at the T-10 tokamak is calculated from the CXRS spectra of the 5291-Å line. The time behavior of the ion temperature profile in different ohmic heating modes is studied. The temperature profile dependence on the ECRH power is measured, and the dynamics of ECR removal of carbon nuclei from the T-10 plasma is described. Experimental data from the CXRS diagnostics at T-10 substantially contribute to the implementation of physical programs of studies on heat and particle transport in tokamak plasmas and investigation of geodesic acoustic mode properties.

Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST

International Nuclear Information System (INIS)

Xu, X Q

2007-01-01

We present gyrokinetic neoclassical simulations of tokamak plasmas with self-consistent electric field for the first time using a fully nonlinear (full-f) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five dimensional computational grid in phase space. The present implementation is a Method of Lines approach where the phase-space derivatives are discretized with finite differences and implicit backwards differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving gyrokinetic Poisson equation with self-consistent poloidal variation. With our 4D (ψ, θ, ε, μ) version of the TEMPEST code we compute radial particle and heat flux, the Geodesic-Acoustic Mode (GAM), and the development of neoclassical electric field, which we compare with neoclassical theory with a Lorentz collision model. The present work provides a numerical scheme and a new capability for self-consistently studying important aspects of neoclassical transport and rotations in toroidal magnetic fusion devices

Impurity seeding in ASDEX upgrade tokamak modeled by COREDIV code

Energy Technology Data Exchange (ETDEWEB)

Galazka, K.; Ivanova-Stanik, I.; Czarnecka, A.; Zagoerski, R. [Institute of Plasma Physics and Laser Microfusion, Warsaw (Poland); Bernert, M.; Kallenbach, A. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Collaboration: ASDEX Upgrade Team

2016-08-15

The self-consistent COREDIV code is used to simulate discharges in a tokamak plasma, especially the influence of impurities during nitrogen and argon seeding on the key plasma parameters. The calculations are performed with and without taking into account the W prompt redeposition in the divertor area and are compared to the experimental results acquired on ASDEX Upgrade tokamak (shots 29254 and 29257). For both impurities the modeling shows a better agreement with the experiment in the case without prompt redeposition. It is attributed to higher average tungsten concentration, which on the other hand seriously exceeds the experimental value. By turning the prompt redeposition process on, the W concentration is lowered, what, in turn, results in underestimation of the radiative power losses. By analyzing the influence of the transport coefficients on the radiative power loss and average W concentration it is concluded that the way to compromise the opposing tendencies is to include the edge-localized mode flushing mechanism into the code, which dominates the experimental particle and energy balance. Also performing the calculations with both anomalous and neoclassical diffusion transport mechanisms included is suggested. (copyright 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH and Co. KGaA Weinheim. This)

User's manual for the Oak Ridge Tokamak Transport Code

International Nuclear Information System (INIS)

Munro, J.K.; Hogan, J.T.; Howe, H.C.; Arnurius, D.E.

1977-02-01

A one-dimensional tokamak transport code is described which simulates a plasma discharge using a fluid model which includes power balances for electrons and ions, conservation of mass, and Maxwell's equations. The modular structure of the code allows a user to add models of various physical processes which can modify the discharge behavior. Such physical processes treated in the version of the code described here include effects of plasma transport, neutral gas transport, impurity diffusion, and neutral beam injection. Each process can be modeled by a parameterized analytic formula or at least one detailed numerical calculation. The program logic of each module is presented, followed by detailed descriptions of each subroutine used by the module. The physics underlying the models is only briefly summarized. The transport code was written in IBM FORTRAN-IV and implemented on IBM 360/370 series computers at the Oak Ridge National Laboratory and on the CDC 7600 computers of the Magnetic Fusion Energy (MFE) Computing Center of the Lawrence Livermore Laboratory. A listing of the current reference version is provided on accompanying microfiche

Conceptual integrated approach for the magnet system of a tokamak reactor

International Nuclear Information System (INIS)

Duchateau, J.-L.; Hertout, P.; Saoutic, B.; Artaud, J.-F.; Zani, L.; Reux, C.

2014-01-01

Highlights: • We give a conceptual approach of a fusion reactor magnet system based on analytical formula. • We give design criteria for the CS and TF cable in conduit conductors and for the magnet system structural description. • We apply this conceptual approach to ITER and we crosscheck with actual characteristics. • We apply this conceptual approach to a possible version of DEMO. - Abstract: In the framework of the reflexion about DEMO, a conceptual integrated approach for the magnet system of a tokamak reactor is presented. This objective is reached using analytical formulas which are presented in this paper, coupled to a Fortran code ESCORT (Electromagnetic Superconducting System for the Computation of Research Tokamaks), to be integrated into SYCOMORE, a code for reactor modelling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modelling task force. The analytical formulas deal with all aspects of the magnet system, starting from the derivation of the TF system general geometry, from the plasma main characteristics. The design criteria for the cable current density and the structural design of the toroidal field and central solenoid systems are presented, enabling to deliver the radial thicknesses of the magnets and enabling also to estimate the plasma duration of the plateau. As a matter of fact, a pulsed version DEMO is presently actively considered in the European programmes. Considerations regarding the cryogenics and the protection are given, affecting the general design. An application of the conceptual approach is presented, allowing a comparison between ESCORT output data and actual ITER parameters and giving the main characteristics of a possible version for DEMO

An advanced computational algorithm for systems analysis of tokamak power plants

International Nuclear Information System (INIS)

Dragojlovic, Zoran; Rene Raffray, A.; Najmabadi, Farrokh; Kessel, Charles; Waganer, Lester; El-Guebaly, Laila; Bromberg, Leslie

2010-01-01

A new computational algorithm for tokamak power plant system analysis is being developed for the ARIES project. The objective of this algorithm is to explore the most influential parameters in the physical, technological and economic trade space related to the developmental transition from experimental facilities to viable commercial power plants. This endeavor is being pursued as a new approach to tokamak systems studies, which examines an expansive, multi-dimensional trade space as opposed to traditional sensitivity analyses about a baseline design point. The new ARIES systems code consists of adaptable modules which are built from a custom-made software toolbox using object-oriented programming. The physics module captures the current tokamak physics knowledge database including modeling of the most-current proposed burning plasma experiment design (FIRE). The engineering model accurately reflects the intent and design detail of the power core elements including accurate and adjustable 3D tokamak geometry and complete modeling of all the power core and ancillary systems. Existing physics and engineering models reflect both near-term as well as advanced technology solutions that have higher performance potential. To fully assess the impact of the range of physics and engineering implementations, the plant cost accounts have been revised to reflect a more functional cost structure, supported by an updated set of costing algorithms for the direct, indirect, and financial cost accounts. All of these features have been validated against the existing ARIES-AT baseline case. The present results demonstrate visualization techniques that provide an insight into trade space assessment of attractive steady-state tokamaks for commercial use.

MHD stability analyses of a tokamak plasma by time-dependent codes

International Nuclear Information System (INIS)

Kurita, Gen-ichi

1982-07-01

The MHD properties of a tokamak plasma are investigated by using time evolutional codes. As for the ideal MHD modes we have analyzed the external modes including the positional instability. Linear and nonlinear ideal MHD codes have been developed. Effects of the toroidicity and conducting shell on the external kink mode are studied minutely by the linear code. A new rezoning algorithm is devised and it is successfully applied to express numerically the axisymmetric plasma perturbation in a cylindrical geometry. As for the resistive MHD modes we have developed nonlinear codes on the basis of the reduced set of the resistive MHD equations. By using the codes we have studied the major disruption processes and properties of the low n resistive modes. We have found that the effects of toroidicity and finite poloidal beta are very important. Considering the above conclusion we propose a new scenario of the initiation of the major disruption. (author)

Conceptual radiation shielding design of superconducting tokamak fusion device by PHITS

International Nuclear Information System (INIS)

Sukegawa, Atsuhiko M.; Kawasaki, Hiromitsu; Okuno, Koichi

2010-01-01

A complete 3D neutron and photon transport analysis by Monte Carlo transport code system PHITS (Particle and Heavy Ion Transport code System) have been performed for superconducting tokamak fusion device such as JT-60 Super Advanced (JT-60SA). It is possible to make use of PHITS in the port streaming analysis around the devices for the tokamak fusion device, the duct streaming analysis in the building where the device is installed, and the sky shine analysis for the site boundary. The neutron transport analysis by PHITS makes it clear that the shielding performance of the superconducting tokamak fusion device with the cryostat is improved by the graphical results. From the standpoint of the port streaming and the duct streaming, it is necessary to calculate by 3D Monte Carlo code such as PHITS for the neutronics analysis of superconducting tokamak fusion device. (author)

Neoclassical simulation of tokamak plasmas using the continuum gyrokinetic code TEMPEST.

Science.gov (United States)

Xu, X Q

2008-07-01

We present gyrokinetic neoclassical simulations of tokamak plasmas with a self-consistent electric field using a fully nonlinear (full- f ) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five-dimensional computational grid in phase space. The present implementation is a method of lines approach where the phase-space derivatives are discretized with finite differences, and implicit backward differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving the gyrokinetic Poisson equation with self-consistent poloidal variation. With a four-dimensional (psi,theta,micro) version of the TEMPEST code, we compute the radial particle and heat fluxes, the geodesic-acoustic mode, and the development of the neoclassical electric field, which we compare with neoclassical theory using a Lorentz collision model. The present work provides a numerical scheme for self-consistently studying important dynamical aspects of neoclassical transport and electric field in toroidal magnetic fusion devices.

Large Aspect Ratio Tokamak Study

International Nuclear Information System (INIS)

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

1980-06-01

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

A study on the fusion reactor - Development of MHD stability and transport code for KT-2 tokamak

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Koo; Shin, Kyo Jin [Pohang University of Science and Tecnology, Pohang (Korea, Republic of)

1996-08-01

MHD Stability analyses for KT-2 Tokamak were carried out by using CART (Resistive 3-D) Code. Linear Growth rates and linear perturbed eigen function of both N=0 axisymmetric mode and N=1 kink modes of highly elongated tokamak plasmas, in the presence of a conducting wall at various distances are computed and linear and nonlinear evolution of N=0 axisymmetric modes are simulated. 26 refs., 25 figs. (author)

Simulation of the MHD stabilities of the experiment on HL-2A tokamak by GATO code

International Nuclear Information System (INIS)

Pan Wei; Chen Liaoyuan; Dong Jiaqi; Shen Yong; Zhang Jinhua

2009-01-01

The ideal two-dimensional MHD stabilities code, GATO, has been successfully immigrated to the high-performance computing system of HL-2A and used to the simulation study of the ideal MHD stabilities of the plasmas produced by one of the pellets injection experiments on HL-2A tokamak. The EFIT code was used to reconstruct the equilibrium configures firstly and the GATO was used to compute their MHD stabilities secondly whose source data were obtained by the NO.4050 discharge of the experiments on HL-2A, and finally by analyzing these results the preliminary conclusion was devised that the confinement performance of the plasma was improved because of the stabilization effect of the anti-sheared configures created by the pellets injection. (authors)

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

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

1985-01-01

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

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

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

1986-01-01

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

User's manual for the Oak Ridge Tokamak Transport Code

Energy Technology Data Exchange (ETDEWEB)

Munro, J.K.; Hogan, J.T.; Howe, H.C.; Arnurius, D.E.

1977-02-01

A one-dimensional tokamak transport code is described which simulates a plasma discharge using a fluid model which includes power balances for electrons and ions, conservation of mass, and Maxwell's equations. The modular structure of the code allows a user to add models of various physical processes which can modify the discharge behavior. Such physical processes treated in the version of the code described here include effects of plasma transport, neutral gas transport, impurity diffusion, and neutral beam injection. Each process can be modeled by a parameterized analytic formula or at least one detailed numerical calculation. The program logic of each module is presented, followed by detailed descriptions of each subroutine used by the module. The physics underlying the models is only briefly summarized. The transport code was written in IBM FORTRAN-IV and implemented on IBM 360/370 series computers at the Oak Ridge National Laboratory and on the CDC 7600 computers of the Magnetic Fusion Energy (MFE) Computing Center of the Lawrence Livermore Laboratory. A listing of the current reference version is provided on accompanying microfiche.

A one-dimensional transport code for the simulation of D-T burning tokamak plasma

International Nuclear Information System (INIS)

Tone, Tatsuzo; Maki, Koichi; Kasai, Masao; Nishida, Hidetsugu

1980-11-01

A one-dimensional transport code for D-T burning tokamak plasma has been developed, which simulates the spatial behavior of fuel ions(D, T), alpha particles, impurities, temperatures of ions and electrons, plasma current, neutrals, heating of alpha and injected beam particles. The basic transport equations are represented by one generalized equation so that the improvement of models and the addition of new equations may be easily made. A model of burn control using a variable toroidal field ripple is employed. This report describes in detail the simulation model, numerical method and the usage of the code. Some typical examples to which the code has been applied are presented. (author)

Combined confinement system applied to tokamaks

International Nuclear Information System (INIS)

Ohkawa, Tihiro

1986-01-01

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

Large aspect ratio tokamak study

International Nuclear Information System (INIS)

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

1979-01-01

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

Design studies of Tokamak power reactor in JAERI

International Nuclear Information System (INIS)

Tone, T.; Nishikawa, M.; Tanaka, Y.

1985-01-01

Recent design studies of tokamak power reactor and related activities conducted in JAERI are presented. A design study of the SPTR (Swimming-Pool Type Reactor) concept was carried out in FY81 and FY82. The reactor design studies in the last two years focus on nuclear components, heat transport and energy conversion systems. In parallel of design studies, tokamak systems analysis code is under development to evaluate reactor performances, cost and net energy balance

Modelling dust transport in tokamaks

International Nuclear Information System (INIS)

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

2008-01-01

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

High performance operational limits of tokamak and helical systems

International Nuclear Information System (INIS)

Yamazaki, Kozo; Kikuchi, Mitsuru

2003-01-01

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

A Toroidally Symmetric Plasma Simulation code for design of position and shape control on tokamak plasmas

International Nuclear Information System (INIS)

Takase, Haruhiko; Senda, Ikuo

1999-01-01

A Toroidally Symmetric Plasma Simulation (TSPS) code has been developed for investigating the position and shape control on tokamak plasmas. The analyses of three-dimensional eddy currents on the conducting components around the plasma and the two-dimensional magneto-hydrodynamic (MHD) equilibrium are taken into account in this code. The code can analyze the plasma position and shape control during the minor disruption in which the deformation of plasma is not negligible. Using the ITER (International Thermonuclear Experimental Reactor) parameters, some examples of calculations are shown in this paper. (author)

Lower hybrid current drive in shaped tokamaks

International Nuclear Information System (INIS)

Kesner, J.

1993-01-01

A time dependent lower hybrid current drive tokamak simulation code has been developed. This code combines the BALDUR tokamak simulation code and the Bonoli/Englade lower hybrid current drive code and permits the study of the interaction of lower hybrid current drive with neutral beam heating in shaped cross-section plasmas. The code is time dependent and includes the beam driven and bootstrap currents in addition to the current driven by the lower hybrid system. Examples of simulations are shown for the PBX-M experiment which include the effect of cross section shaping on current drive, ballooning mode stabilization by current profile control and sawtooth stabilization. A critical question in current drive calculations is the radial transport of the energetic electrons. The authors have developed a response function technique to calculate radial transport in the presence of an electric field. The consequences of the combined influences of radial diffusion and electric field acceleration are discussed

A ''SuperCode'' for performing systems analysis of tokamak experiments and reactors

International Nuclear Information System (INIS)

Haney, S.W.; Barr, W.L.; Crotinger, J.A.; Perkins, L.J.; Solomon, C.J.; Chaniotakis, E.A.; Freidberg, J.P.; Wei, J.; Galambos, J.D.; Mandrekas, J.

1992-01-01

A new code, named the ''SUPERCODE,'' has been developed to fill the gap between currently available zero dimensional systems codes and highly sophisticated, multidimensional plasma performance codes. The former are comprehensive in content, fast to execute, but rather simple in terms of the accuracy of the physics and engineering models. The latter contain state-of-the-art plasma physics modelling but are limited in engineering content and time consuming to run. The SUPERCODE upgrades the reliability and accuracy of systems codes by calculating the self consistent 1 1/2 dimensional MHD-transport plasma evolution in a realistic engineering environment. By a combination of variational techniques and careful formation, there is only a modest increase in CPU time over O-D runs, thereby making the SUPERCODE suitable for use as a systems studies tool. In addition, considerable effort has been expended to make the code user- and programming-friendly, as well as operationally flexible, with the hope of encouraging wide usage throughout the fusion community

FLOC: Field Line and Orbit Code for the study of ripple beam injection into tokamaks

Energy Technology Data Exchange (ETDEWEB)

Fowler, R. H.; Lee, D. K.; Gaffney, P. W.; Rome, J. A.

1978-06-01

The computer code described is used to study ripple beam injection into a tokamak plasma. The collisionless guiding center equations of motion are integrated to find the orbits of single particles in realistic magnetic fields for ripple injection. In order to determine if the ripple is detrimental to the plasma, the magnetic flux surfaces are constructed by integration of the field line equations. The numerical techniques are described, and use of the code is outlined. A program listing is provided, and the results of sample cases are presented.

FLOC: Field Line and Orbit Code for the study of ripple beam injection into tokamaks

International Nuclear Information System (INIS)

Fowler, R.H.; Lee, D.K.; Gaffney, P.W.; Rome, J.A.

1978-06-01

The computer code described is used to study ripple beam injection into a tokamak plasma. The collisionless guiding center equations of motion are integrated to find the orbits of single particles in realistic magnetic fields for ripple injection. In order to determine if the ripple is detrimental to the plasma, the magnetic flux surfaces are constructed by integration of the field line equations. The numerical techniques are described, and use of the code is outlined. A program listing is provided, and the results of sample cases are presented

Parametric system studies of candidate TF coil system options for the Tokamak Fusion Core Experiment (TFCX)

International Nuclear Information System (INIS)

Reiersen, W.T.; Flanagan, C.A.; Miller, J.B.

1983-01-01

System studies were performed to determine the sensitivity of hybrid and superconducting toroidal field (TF) coil system options to maximum field at the TF coil and to field enhancement due to resistive insert coils. The studies were performed using Tokamak Fusion Core Experiment (TFCX) design assumptions, guidelines, and criteria and involved iterative execution of the Fusion Engineering Design Center (FEDC) systems code, magnetohydrodynamics (MHD) equilibrium code, and EFFI (a code to evaluate magnetic field strength). The results indicate that for TFCX with no minimum wall loading specified, a design point chosen solely on the basis of cost would likely be in the low-field region of design space where the cost advantage of hybrids is least apparent. However, as the desired neutron wall loading increases, the hybrid option suggests an increasing cost advantage over the all-superconducting option; this cost advantage is countered by increased complexity in design -- particularly in assembly and maintenance

Parametric system studies of candidate TF coil system options for the Tokamak Fusion Core Experiment (TFCX)

International Nuclear Information System (INIS)

Reiersen, W.T.; Flanagan, C.A.; Miller, J.B.

1983-01-01

System studies were performed to determine the sensitivity of hybrid and superconducting toroidal field (TF) coil system options to maximum field at the TF coil and to field enhancement due to resistive insert coils. The studies were performed using Tokamak Fusion Core Experiment (TFCX) design assumptions, guidelines, and criteria and involved iterative execution of the Fusion Engineering Design Center (FEDC) systems code, magnetohydrodynamics (MHD) equilibrium code, and EFFI (a code to evaluate magnetic field strength). The results indicate that for TFCX with no minimum wall loading specified, a design point chosen solely on the basis of cost would likely be in the low-field region of design space where the cost advantage of hybrids is least apparent. However, as the desired neutron wall loading increases, the hybrid option suggests an increasing cost advantage over the all-superconducting option; this cost advantage is countered by increased complexity in design - particularly in assembly and maintenance

Computational studies of tokamak plasmas

International Nuclear Information System (INIS)

Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

1981-02-01

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

HELIAS module development for systems codes

Energy Technology Data Exchange (ETDEWEB)

Warmer, F., E-mail: Felix.Warmer@ipp.mpg.de; Beidler, C.D.; Dinklage, A.; Egorov, K.; Feng, Y.; Geiger, J.; Schauer, F.; Turkin, Y.; Wolf, R.; Xanthopoulos, P.

2015-02-15

In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations. This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work. This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.

CASINO, a code for simulation of charged particles in an axisymmetric Tokamak

International Nuclear Information System (INIS)

Dillner, Oe.

1992-01-01

The present report comprises a documentation of CASINO, a simulation code developed as a means for the study of high energy charged particles in an axisymmetric Tokamak. The background of the need for such a numerical tool is presented. In the description of the numerical model used for the orbit integration, the method using constants of motion, the Lao-Hirsman geometry for the flux surfaces and a method for reducing the necessary number of particles is elucidated. A brief outline of the calculational sequence is given as a flow chart. The essential routines and functions as well as the common blocks are briefly described. The input and output routines are shown. Finally the documentation is completed by a short discussion of possible extensions of the code and a test case. (au)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Physics parameter space of tokamak ignition devices

International Nuclear Information System (INIS)

Selcow, E.C.; Peng, Y.K.M.; Uckan, N.A.; Houlberg, W.A.

1985-01-01

This paper describes the results of a study to explore the physics parameter space of tokamak ignition experiments. A new physics systems code has been developed to perform the study. This code performs a global plasma analysis using steady-state, two-fluid, energy-transport models. In this paper, we discuss the models used in the code and their application to the analysis of compact ignition experiments. 8 refs., 8 figs., 1 tab

Electronic system of TBR tokamak device

International Nuclear Information System (INIS)

Silva, R.P. da.

1980-01-01

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

Conceptual Design of Alborz Tokamak Poloidal Coils System

Science.gov (United States)

Mardani, M.; Amrollahi, R.

2013-04-01

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

User's guide for SLWDN9, a code for calculating flux-surfaced-averaging of alpha densities, currents, and heating in non-circular tokamaks

International Nuclear Information System (INIS)

Hively, L.M.; Miley, G.M.

1980-03-01

The code calculates flux-surfaced-averaged values of alpha density, current, and electron/ion heating profiles in realistic, non-circular tokamak plasmas. The code is written in FORTRAN and execute on the CRAY-1 machine at the Magnetic Fusion Energy Computer Center

Neutronic analysis of fusion tokamak devices by PHITS

International Nuclear Information System (INIS)

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

2011-01-01

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

Multi-mode remote participation on the GOLEM tokamak

International Nuclear Information System (INIS)

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

2011-01-01

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

A flexible software architecture for tokamak discharge control systems

International Nuclear Information System (INIS)

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

1995-01-01

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

Hybrid parallelization of the XTOR-2F code for the simulation of two-fluid MHD instabilities in tokamaks

Science.gov (United States)

Marx, Alain; Lütjens, Hinrich

2017-03-01

A hybrid MPI/OpenMP parallel version of the XTOR-2F code [Lütjens and Luciani, J. Comput. Phys. 229 (2010) 8130] solving the two-fluid MHD equations in full tokamak geometry by means of an iterative Newton-Krylov matrix-free method has been developed. The present work shows that the code has been parallelized significantly despite the numerical profile of the problem solved by XTOR-2F, i.e. a discretization with pseudo-spectral representations in all angular directions, the stiffness of the two-fluid stability problem in tokamaks, and the use of a direct LU decomposition to invert the physical pre-conditioner at every Krylov iteration of the solver. The execution time of the parallelized version is an order of magnitude smaller than the sequential one for low resolution cases, with an increasing speedup when the discretization mesh is refined. Moreover, it allows to perform simulations with higher resolutions, previously forbidden because of memory limitations.

Diagnostics systems for the TBR-E tokamak

International Nuclear Information System (INIS)

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

1992-08-01

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

Stability analysis of ELMs in long-pulse discharges with ELITE code on EAST tokamak

Science.gov (United States)

Wang, Y. F.; Xu, G. S.; Wan, B. N.; Li, G. Q.; Yan, N.; Li, Y. L.; Wang, H. Q.; Peng, Y.-K. Martin; Xia, T. Y.; Ding, S. Y.; Chen, R.; Yang, Q. Q.; Liu, H. Q.; Zang, Q.; Zhang, T.; Lyu, B.; Xu, J. C.; Feng, W.; Wang, L.; Chen, Y. J.; Luo, Z. P.; Hu, G. H.; Zhang, W.; Shao, L. M.; Ye, Y.; Lan, H.; Chen, L.; Li, J.; Zhao, N.; Wang, Q.; Snyder, P. B.; Liang, Y.; Qian, J. P.; Gong, X. Z.; EAST team

2018-05-01

One challenge in long-pulse and high performance tokamak operation is to control the edge localized modes (ELMs) to reduce the transient heat load on plasma facing components. Minute-scale discharges in H-mode have been achieved repeatedly on Experimental Advanced Superconducting Tokamak (EAST) since the 2016 campaign and understanding the characteristics of the ELMs in these discharges can be helpful for effective ELM control in long-pulse discharges. The kinetic profile diagnostics recently developed on EAST make it possible to perform the pedestal stability analysis quantitatively. Pedestal stability calculation of a typical long-pulse discharge with ELITE code is presented. The ideal linear stability results show that the ELM is dominated by toroidal mode number n around 10–15 and the most unstable mode structure is mainly localized in the steep pressure gradient region, which is consistent with experimental results. Compared with a typical type-I ELM discharge with larger total plasma current (I p = 600 kA), pedestal in the long-pulse H-mode discharge (I p = 450 kA) is more stable in peeling-ballooning instability and its critical peak pressure gradient is evaluated to be 65% of the former. Two important features of EAST tokamak in the long-pulse discharge are presented by comparison with other tokamaks, including a wider pedestal correlated with the poloidal pedestal beta and a smaller inverse aspect ratio and their effects on the pedestal stability are discussed. The effects of uncertainties in measurements on the linear stability results are also analyzed, including the edge electron density profile position, the separatrix position and the line-averaged effective ion charge {Z}{{e}{{f}}{{f}}} value.

Comparative study of cost models for tokamak DEMO fusion reactors

International Nuclear Information System (INIS)

Oishi, Tetsutarou; Yamazaki, Kozo; Arimoto, Hideki; Ban, Kanae; Kondo, Takuya; Tobita, Kenji; Goto, Takuya

2012-01-01

Cost evaluation analysis of the tokamak-type demonstration reactor DEMO using the PEC (physics-engineering-cost) system code is underway to establish a cost evaluation model for the DEMO reactor design. As a reference case, a DEMO reactor with reference to the SSTR (steady state tokamak reactor) was designed using PEC code. The calculated total capital cost was in the same order of that proposed previously in cost evaluation studies for the SSTR. Design parameter scanning analysis and multi regression analysis illustrated the effect of parameters on the total capital cost. The capital cost was predicted to be inside the range of several thousands of M$s in this study. (author)

Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code

International Nuclear Information System (INIS)

Evans, T.E.; Mahdavi, M.A.; Sager, G.T.; West, W.P.; Fenstermacher, M.E.; Meyer, W.H.; Porter, G.D.

1995-07-01

A Monte-Carlo Impurity (MCI) transport code is used to follow trace impurities through multiple ionization states in realistic 2-D tokamak geometries. The MCI code is used to study impurity transport along the open magnetic field lines of the Scrape-off Layer (SOL) and to understand how impurities get into the core from the SOL. An MCI study concentrating on the entrainment of carbon impurities ions by deuterium background plasma into the DIII-D divertor is discussed. MCI simulation results are compared to experimental DIII-D carbon measurements

Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code

International Nuclear Information System (INIS)

Evans, T.E.; Sager, G.T.; Mahdavi, M.A.; Porter, G.D.; Fenstermacher, M.E.; Meyer, W.H.

1995-01-01

A Monte-Carlo Impurity (MCI) transport code is used to follow trace impurities through multiple ionization states in realistic 2-D tokamak geometries. The MCI code is used to study impurity transport along the open magnetic field lines of the Scrape-off Layer (SOL) and to understand how impurities get into the core from the SOL. An MCI study concentrating on the entrainment of carbon impurities ions by deuterium background plasma into the DII-D divertor is discussed. MCI simulation results are compared to experimental DII-D carbon measurements. 2 refs

Cryogenic system design for a compact tokamak reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.; Miller, J.R.

1988-01-01

The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

Compact Commercial Tokamak Reactor (CCTR): a concept for a 500-MWe commercial-tokamak fusion system

International Nuclear Information System (INIS)

Gillen, T.J.

1980-11-01

A detailed set of self-consistent parameters and costs for the conceptual design of a Compact Commercial Tokamak Reactor (CCTR) is given. Several of the basic design features are the following: an ignited plasma with a major radius of 4.9 m and minor radius of 1.4 m; a net electrical output of 500 MW; a borated-water-cooled, stainless steel shield; and a toroidal field of 12 T at the coil. The design, which utilizes the Westinghouse computer code for the COsting And Sizing of D-T burning Tokamaks (COAST), mainly provides the sizes and geometries associated with the definition of the main component features for which a detailed engineering design can be effectively undertaken. Design study alternatives, including a neutral beam driven design option, a design option with a toroidal field of 13 T at the coil, and a tungsten-shielded option are considered for the CCTR. Also included is the conceptual design of a Compact Fusion Engineering Device

Continuous, saturation, and discontinuous tokamak plasma vertical position control systems

Energy Technology Data Exchange (ETDEWEB)

Mitrishkin, Yuri V., E-mail: y_mitrishkin@hotmail.com [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Pavlova, Evgeniia A., E-mail: janerigoler@mail.ru [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Kuznetsov, Evgenii A., E-mail: ea.kuznetsov@mail.ru [Troitsk Institute for Innovation and Fusion Research, Moscow 142190 (Russian Federation); Gaydamaka, Kirill I., E-mail: k.gaydamaka@gmail.com [V. A. Trapeznikov Institute of Control Sciences of the Russian Academy of Sciences, Moscow 117997 (Russian Federation)

2016-10-15

Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.

Continuous, saturation, and discontinuous tokamak plasma vertical position control systems

International Nuclear Information System (INIS)

Mitrishkin, Yuri V.; Pavlova, Evgeniia A.; Kuznetsov, Evgenii A.; Gaydamaka, Kirill I.

2016-01-01

Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.

Tokamak power systems studies, FY 1985

International Nuclear Information System (INIS)

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

1985-12-01

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

Tokamak power systems studies, FY 1985

Energy Technology Data Exchange (ETDEWEB)

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

1985-12-01

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

Conceptual design of Remote Control System for EAST tokamak

International Nuclear Information System (INIS)

Sun, X.Y.; Wang, F.; Wang, Y.; Li, S.

2014-01-01

Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication

Conceptual design of Remote Control System for EAST tokamak

Energy Technology Data Exchange (ETDEWEB)

Sun, X.Y., E-mail: xysun@ipp.ac.cn; Wang, F.; Wang, Y.; Li, S.

2014-05-15

Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.

Reconnection in tokamaks

International Nuclear Information System (INIS)

Pare, V.K.

1983-01-01

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

Data processing system for spectroscopy at Novillo Tokamak

International Nuclear Information System (INIS)

Ortega C, G.; Gaytan G, E.

1998-01-01

Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

Transmutation blanket design for a Tokamak system

International Nuclear Information System (INIS)

Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria A. Fortini; Costa, Antonella L.

2011-01-01

Sub-critical advanced reactor with a D-T fusion neutron source based on Tokamak technology is an innovative type of nuclear system. Due to the high quantity of neutrons produced by fusion reactions, it could be well spent in the transmutation process of the transuranic elements. Nevertheless, to achieve a successful transmutation, it is necessary to know the neutron fluence along the radial axis and its characteristics. In this work, it evaluated the neutron flux and interaction frequency along the radial axis changing the material of the first wall. W-alloy, beryllium and the combination of both were studied and regions more suitable to transmutation were determined. The results demonstrated that the better zone to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements W-alloy/W-alloy and W-alloy/Beryllium would be able to hold the requirements of high fluence and hardening spectrum needed to transuranic transmutation. The system was simulated using the MCNP5 code, the ITER Final Design Report, 2001, and the FENDL/MC-2.1 nuclear data library. (author)

System assessment of helical reactors in comparison with tokamaks

International Nuclear Information System (INIS)

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

2002-10-01

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

Tokamak poloidal-field systems. Progress report, January 1-December 31, 1982

International Nuclear Information System (INIS)

Rogers, J.D.

1983-05-01

The work performed in support of the FED and INTOR tokamak studies is reported at length and covers almost all the aspects of poloidal field (PF) design that were considered. The design work included magnetics, forces and fields, superconductor design, superconductor loss calculations, high field tokamak central solenoid parametric analysis, helium vapor release with bubble clearing and entrainment analysis, eddy current losses in dewars, structural support design for internally cooled cable superconductor (ICCS), research and technology development and manufacturing plans and milestones for poloidal field (PF) coils, fault conditions for shorted PF coils, design of 50 kA vapor cooled leads, and structural design of PF ring coils box frame dewars. Eddy current calculations in tokamak structure are being calculated. A computer code to perform stability analysis of ICCS is being written. Two water cooled switches, a vacuum interrupter and a bypass switch, were tested to develop improved higher current carrying capacities

A Study of Performance in Low-Power Tokamak Reactor with Integrated Predictive Modeling Code

International Nuclear Information System (INIS)

Pianroj, Y.; Onjun, T.; Suwanna, S.; Picha, R.; Poolyarat, N.

2009-07-01

Full text: A fusion hybrid or a small fusion power output with low power tokamak reactor is presented as another useful application of nuclear fusion. Such tokamak can be used for fuel breeding, high-level waste transmutation, hydrogen production at high temperature, and testing of nuclear fusion technology components. In this work, an investigation of the plasma performance in a small fusion power output design is carried out using the BALDUR predictive integrated modeling code. The simulations of the plasma performance in this design are carried out using the empirical-based Mixed Bohm/gyro Bohm (B/gB) model, whereas the pedestal temperature model is based on magnetic and flow shear (δ α ρ ζ 2 ) stabilization pedestal width scaling. The preliminary results using this core transport model show that the central ion and electron temperatures are rather pessimistic. To improve the performance, the optimization approach are carried out by varying some parameters, such as plasma current and power auxiliary heating, which results in some improvement of plasma performance

Distributed digital real-time control system for TCV tokamak

Energy Technology Data Exchange (ETDEWEB)

Le, H.B. [École Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association EURATOM-Confédération Suisse, CH-1015 Lausanne (Switzerland); Felici, F. [Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands); Paley, J.I.; Duval, B.P.; Moret, J.-M.; Coda, S.; Sauter, O.; Fasel, D.; Marmillod, P. [École Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association EURATOM-Confédération Suisse, CH-1015 Lausanne (Switzerland)

2014-03-15

Highlights: • A new distributed digital control system for the TCV tokamak has been commissioned. • Data is shared in real-time between all nodes using the reflective memory. • The customised Linux OS allows achieving deterministic and low latency behaviour. • The control algorithm design in Simulink together with the automatic code generation using Embedded Coder allow rapid algorithm development. • Controllers designed outside the TCV environment can be ported easily. • The previous control system functions have been emulated and improved. • New capabilities include MHD control, profile control, equilibrium reconstruction. - Abstract: A new digital feedback control system (named the SCD “Système de Contrôle Distribué”) has been developed, integrated and used successfully to control TCV (Tokamak à Configuration Variable) plasmas. The system is designed to be modular, distributed, and scalable, accommodating hundreds of diagnostic inputs and actuator outputs. With many more inputs and outputs available than previously possible, it offers the possibility to design advanced control algorithms with better knowledge of the plasma state and to coherently control all TCV actuators, including poloidal field (PF) coils, gas valves, the gyrotron powers and launcher angles of the electron cyclotron heating and current drive system (ECRH/ECCD) together with diagnostic triggering signals. The system consists of multiple nodes; each is a customised Linux desktop or embedded PC which may have local ADC and DAC cards. Each node is also connected to a memory network (reflective memory) providing a reliable, deterministic method of sharing memory between all nodes. Control algorithms are programmed as block diagrams in Matlab-Simulink providing a powerful environment for modelling and control design. The C code is generated automatically from the Simulink block diagram and compiled, with the Simulink Embedded Coder (SEC, formerly Real-Time Workshop Embedded

Ballooning stable high beta tokamak equilibria

International Nuclear Information System (INIS)

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

1981-04-01

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

The magnet system of the Tokamak T-15 upgrade

International Nuclear Information System (INIS)

Khvostenko, P.P.; Azizov, E.A.; Alfimov, D.E.; Belyakov, V.A.; Bondarchuk, E.N.; Chudnovsky, A.N.; Dokuka, V.N.; Kavin, A.A.; Khayrutdinov, R.R.; Khokhlov, M.V.; Kitaev, B.A.; Krasnov, S.V.; Maximova, I.I.; Labusov, A.N.; Lukash, V.E.; Mineev, A.B.; Muratov, V.P.

2015-01-01

Highlights: • T-15U project is the initial technical base for creating fusion neutron sources. • Magnet system of T-15U will confine the hot plasma in the divertor configuration. • Toroidal magnetic field at the plasma axis is 2 T. • T-15U should begin operations in 2016. - Abstract: Presently, the Tokamak T-15 is being upgraded. The magnet system of the Tokamak T-15 upgrade will obtain and confine the hot plasma in the divertor configuration. Plasma parameters are a major radius of 1.48 m, a minor radius of 0.67 m, an elongation of 1.7–1.9 and a triangularity of 0.3–0.4. The magnet system includes the toroidal winding and the poloidal magnet system. The poloidal magnet system generates the divertor with single null and double null magnetic configurations. The power supply system provides the necessary current scenarios in the windings of the magnet system. All elements of the magnet system will be manufactured by the end of 2015. The Tokamak T-15 upgrade should begin operations in 2016.

The magnet system of the Tokamak T-15 upgrade

Energy Technology Data Exchange (ETDEWEB)

Khvostenko, P.P., E-mail: ppkhvost@rambler.ru [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Azizov, E.A.; Alfimov, D.E. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Belyakov, V.A.; Bondarchuk, E.N. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Chudnovsky, A.N.; Dokuka, V.N. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Kavin, A.A. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Khayrutdinov, R.R. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Khokhlov, M.V.; Kitaev, B.A.; Krasnov, S.V.; Maximova, I.I.; Labusov, A.N. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Lukash, V.E. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Mineev, A.B.; Muratov, V.P. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); and others

2015-10-15

Highlights: • T-15U project is the initial technical base for creating fusion neutron sources. • Magnet system of T-15U will confine the hot plasma in the divertor configuration. • Toroidal magnetic field at the plasma axis is 2 T. • T-15U should begin operations in 2016. - Abstract: Presently, the Tokamak T-15 is being upgraded. The magnet system of the Tokamak T-15 upgrade will obtain and confine the hot plasma in the divertor configuration. Plasma parameters are a major radius of 1.48 m, a minor radius of 0.67 m, an elongation of 1.7–1.9 and a triangularity of 0.3–0.4. The magnet system includes the toroidal winding and the poloidal magnet system. The poloidal magnet system generates the divertor with single null and double null magnetic configurations. The power supply system provides the necessary current scenarios in the windings of the magnet system. All elements of the magnet system will be manufactured by the end of 2015. The Tokamak T-15 upgrade should begin operations in 2016.

A Fast Shutdown Technique for Large Tokamaks

International Nuclear Information System (INIS)

Fredrickson, E.; Schmidt, G.L.; Hill, K.; Jardin, S.C.

1999-01-01

A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-50 deuterium pellets doped with a small ( 0.0005%) concentration of Krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modeling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR

Spherical tokamak power plant design issues

International Nuclear Information System (INIS)

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

2000-01-01

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

Time-dependent free boundary equilibrium and resistive diffusion in a tokamak plasma

International Nuclear Information System (INIS)

Selig, G.

2012-12-01

In a Tokamak, in order to create the necessary conditions for nuclear fusion to occur, a plasma is maintained by applying magnetic fields. Under the hypothesis of an axial symmetry of the tokamak, the study of the magnetic configuration at equilibrium is done in two dimensions, and is deduced from the poloidal flux function. This function is solution of a non linear partial differential equation system, known as equilibrium problem. This thesis presents the time dependent free boundary equilibrium problem, where the circuit equations in the tokamak coils and passive conductors are solved together with the Grad-Shafranov equation to produce a dynamic simulation of the plasma. In this framework, the Finite Element equilibrium code CEDRES has been improved in order to solve the aforementioned dynamic problem. Consistency tests and comparisons with the DINA-CH code on an ITER vertical instability case have validated the results. Then, the resistive diffusion of the plasma current density has been simulated using a coupling between CEDRES and the averaged one-dimensional diffusion equation, and it has been successfully compared with the integrated modeling code CRONOS. (author)

3D simulation studies of tokamak plasmas using MHD and extended-MHD models

International Nuclear Information System (INIS)

Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.

1996-01-01

The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω *i stabilization and nonlinear island saturation of TAE mode using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D ++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree which agree well with experimental data

Advanced commercial Tokamak optimization studies

International Nuclear Information System (INIS)

Whitley, R.H.; Berwald, D.H.; Gordon, J.D.

1985-01-01

Our recent studies have concentrated on developing optimal high beta (bean-shaped plasma) commercial tokamak configurations using TRW's Tokamak Reactor Systems Code (TRSC) with special emphasis on lower net electric power reactors that are more easily deployable. A wide range of issues were investigated in the search for the most economic configuration: fusion power, reactor size, wall load, magnet type, inboard blanket and shield thickness, plasma aspect ratio, and operational β value. The costs and configurations of both steady-state and pulsed reactors were also investigated. Optimal small and large reactor concepts were developed and compared by studying the cost of electricity from single units and from multiplexed units. Multiplexed units appear to have advantages because they share some plant equipment and have lower initial capital investment as compared to larger single units

Modular tokamak magnetic system

International Nuclear Information System (INIS)

Yang, T.F.

1988-01-01

This patent describes a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of the vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein the magnetic coil is wound around the toroidal field coil such that the ohmic heating coil enclosed the toroidal field coil

Proposed tokamak poloidal field system development program

Energy Technology Data Exchange (ETDEWEB)

Rogers, J.D.; Vogel, H.F.; Warren, R.W.; Weldon, D.M.

1977-05-01

A program is proposed to develop poloidal field components for TNS and EPR size tokamak devices and to test these components in realistic circuits. Emphasis is placed upon the development of the most difficult component, the superconducting ohmic-heating coil. Switches must also be developed for testing the coils, and this switching technology is to be extended to meet the requirements for the large scale tokamaks. Test facilities are discussed; power supplies, including a homopolar to drive the coils, are considered; and poloidal field systems studies are proposed.

ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

International Nuclear Information System (INIS)

HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG

2002-01-01

A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response

3D simulation studies of tokamak plasmas using MHD and extended-MHD models

International Nuclear Information System (INIS)

Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.; Pomphrey, N.; Sugiyama, L.E.

1997-01-01

The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω *i stabilization and nonlinear island rotation studies using the two-fluid level MH3D-T code, studies of nonlinear saturation of TAE modes using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D ++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree well with experimental data

New applications of Equinox code for real-time plasma equilibrium and profile reconstruction for tokamaks

International Nuclear Information System (INIS)

Bosak, K.; Blum, J.; Joffrin, E.

2004-01-01

Recent development of real-time equilibrium code Equinox using a fixed-point algorithm allow major plasma magnetic parameters to be identified in real-time, using rigorous analytical method. The code relies on the boundary flux code providing magnetic flux values on the first wall of vacuum vessel. By means of least-square minimization of differences between magnetic field obtained from previous solution and the next measurements the code identifies the source term of the non-linear Grad-Shafranov equation. The strict use of analytical equations together with a flexible algorithm offers an opportunity to include new measurements into stable magnetic equilibrium code and compare the results directly between several tokamaks while maintaining the same physical model (i.e. no iron model is necessary inside the equilibrium code). The successful implementation of this equilibrium code for JET and Tore Supra has already been published. In this paper, we show the preliminary results of predictive runs of the Equinox code using the ITER geometry. Because the real-time control experiments of plasma profile at JET using the code has been shown unstable when using magnetic and polarimetric measurements (that could be indirectly translated into accuracy vs robustness tradeoff), we plan an outline of the algorithm that will allow us to further constrain the plasma current profile using the central value of pressure of the plasma in real-time in order to better define the poloidal beta (this constraint is not necessary with purely magnetic equilibrium). (authors)

New applications of Equinox code for real-time plasma equilibrium and profile reconstruction for tokamaks

Energy Technology Data Exchange (ETDEWEB)

Bosak, K.; Blum, J. [Universite de Nice-Sophia-Antipolis, Lab. J. A. Dieudonne, 06 - Nice (France); Joffrin, E. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

2004-07-01

Recent development of real-time equilibrium code Equinox using a fixed-point algorithm allow major plasma magnetic parameters to be identified in real-time, using rigorous analytical method. The code relies on the boundary flux code providing magnetic flux values on the first wall of vacuum vessel. By means of least-square minimization of differences between magnetic field obtained from previous solution and the next measurements the code identifies the source term of the non-linear Grad-Shafranov equation. The strict use of analytical equations together with a flexible algorithm offers an opportunity to include new measurements into stable magnetic equilibrium code and compare the results directly between several tokamaks while maintaining the same physical model (i.e. no iron model is necessary inside the equilibrium code). The successful implementation of this equilibrium code for JET and Tore Supra has already been published. In this paper, we show the preliminary results of predictive runs of the Equinox code using the ITER geometry. Because the real-time control experiments of plasma profile at JET using the code has been shown unstable when using magnetic and polarimetric measurements (that could be indirectly translated into accuracy vs robustness tradeoff), we plan an outline of the algorithm that will allow us to further constrain the plasma current profile using the central value of pressure of the plasma in real-time in order to better define the poloidal beta (this constraint is not necessary with purely magnetic equilibrium). (authors)

Code of a Tokamak Fusion Energy Facility ITER

International Nuclear Information System (INIS)

Yasuhide Asada; Kenzo Miya; Kazuhiko Hada; Eisuke Tada

2002-01-01

The technical structural code for ITER (International Thermonuclear Experimental Fusion Reactor) and, as more generic applications, for D-T burning fusion power facilities (hereafter, Fusion Code) should be innovative because of their quite different features of safety and mechanical components from nuclear fission reactors, and the necessity of introducing several new fabrication and examination technologies. Introduction of such newly developed technologies as inspection-free automatic welding into the Fusion Code is rationalized by a pilot application of a new code concept of s ystem-based code for integrity . The code concept means an integration of element technical items necessary for construction, operation and maintenance of mechanical components of fusion power facilities into a single system to attain an optimization of the total margin of these components. Unique and innovative items of the Fusion Code are typically as follows: - Use of non-metals; - Cryogenic application; - New design margins on allowable stresses, and other new design rules; - Use of inspection-free automatic welding, and other newly developed fabrication technologies; - Graded approach of quality assurance standard to cover radiological safety-system components as well as non-safety-system components; - Consideration on replacement components. (authors)

Turbulence and abnormal transport in tokamak plasmas

International Nuclear Information System (INIS)

Garbet, X.

1988-09-01

Microinstabilities in linear and nonlinear tokamak plasmas were studied. A variational method based on the existence of a system of angular variables and action for the charged particles in the magnetic configuration of a tokamak is described. The corresponding functional, extremal in relation to the fluctuating electromagnetic field, is calculated analytically, taking into account the effects of the toroidal geometry. A numerical code, TORRID, was derived from these principles and the main instabilities, especially ion instabilities and microtearing, were studied linearly. Nonlinear methods were also applied to microtearing. Quasi-linear transport coefficients are derived from a principle of minimum entropy production. Thermal ionic conductivity and viscosity are calculated for an ionic turbulence [fr

A review of lifetime analyses for tokamaks

International Nuclear Information System (INIS)

Harkness, S.D.; Cramer, B.

1979-01-01

System studies have shown that economic fusion power can best be achieved from the use of long lived components. The stresses generated in a first wall module are a complex function of its geometry, the chosen structural material and the tokamak burn cycle characteristics. A means of applying ASME Code Case 1592 to preliminary design has been established. Methods of incorporating some of the material property changes expected from irradiation are discussed. Cyclic stresses imposed by tokamak operation are expected to cause fatigue related properties to govern the life of the structure. Stress assisted bubble growth is also discussed. This may be the critical mechanism in establishing the creep rupture life of a fusion first wall component. (orig.)

Magnetic ''islandography'' in tokamaks

International Nuclear Information System (INIS)

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

1978-09-01

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

Development of lab scale fast gas injection system for SST-1 Tokamak

International Nuclear Information System (INIS)

Pathan, F.S.; Banaudha, Moni; Khristi, Yohan; Khan, M.S.; Khan, Ziauddin; Raval, D.C.; Khirwadkar, Samir

2017-01-01

The plasma density control plays an important role in Tokamak operation. The factors that influence plasma density in a Tokamak device are working gas injection, pumping, ionization rate and the recycle coefficient representing the wall conditions. Among these factors, gas injection is relatively convenient to be controlled. Hence, the most frequently adopted method to control the plasma density is to control the fast gas injection. This paper describes the design and experimental work carried out towards the development of Fast Gas Injection System for SST-1 Tokamak. Laboratory based test setup was successfully established for Fast Gas Injection System that can feed predefined quantity of gas in a controlled manner into vacuum chamber. Further, this FGIS system will be implemented in SST-1 Tokamak environment with online density feedback signal

Overview of the ITER Tokamak complex building and integration of plant systems toward construction

Energy Technology Data Exchange (ETDEWEB)

Cordier, Jean-Jacques, E-mail: jean-jacques.cordier@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bak, Joo-Shik [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Baudry, Alain [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Benchikhoune, Magali [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Carafa, Leontin; Chiocchio, Stefano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Darbour, Romaric [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Elbez, Joelle; Di Giuseppe, Giovanni; Iwata, Yasuhiro; Jeannoutot, Thomas; Kotamaki, Miikka; Kuehn, Ingo; Lee, Andreas; Levesy, Bruno; Orlandi, Sergio [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Packer, Rachel [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Patisson, Laurent; Reich, Jens; Rigoni, Giuliano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

2015-10-15

The ITER Tokamak complex consists of Tokamak, diagnostic and tritium buildings. The Tokamak machine is located in the bioshield pit of the Tokamak building. Plant systems are implemented in the three buildings and are strongly interfacing with the Tokamak. The reference baseline (3D) configuration is a set of over 1000 models that today defines in an exhaustive way the overall layout of Tokamak and plant systems, needed for fixing the interfaces and to complete the construction design of the buildings. During the last two years, one of the main ITER challenges was to improve the maturity of the plant systems layout in order to confirm their integration in the building final design and freeze the interface definitions in-between the systems and to the buildings. The propagation of safety requirements in the design of the nuclear building like confinement, fire zoning and radiation shielding is of first priority. A major effort was placed by ITER Organization together with the European Domestic Agency (F4E) and the Architect Engineer as a joint team to fix the interfaces and the loading conditions to buildings. The most demanding systems in terms of interface definition are water cooling, cryogenic, detritiation, vacuum, cable trays and building services. All penetrations through the walls for piping, cables and other equipment have been defined, as well as all temporary openings needed for the installation phase. Project change requests (PCR) impacting the Tokamak complex buildings have been implemented in a tight allocated time schedule. The most demanding change was to implement a new design of the Tokamak basic machine supporting system. The 18 supporting columns of the cryostat (2001 baseline) were replaced at the end of 2012 by a concrete crown and radial concrete ribs linked to the basemat and to the bioshield surrounding the Tokamak. The change was implemented successfully in the building construction design to allow basemat construction phase being performed

Design and realization of the J-TEXT tokamak central control system

International Nuclear Information System (INIS)

Yang Zhoujun; Zhuang Ge; Hu Xiwei; Zhang Ming; Qiu Shengshun; Wang Zhijiang; Ding Yonghua; Pan Yuan

2009-01-01

The Joint Texas Experimental Tokamak (J-TEXT), a medium-sized conventional tokamak, serves as a user experimental facility in the China-USA fusion research community. Development of a flexible and easy-to-use J-TEXT central control system (CCS) is of supreme importance for users to coordinate the experimental scenarios with full integration into the discharge operation. This paper describes in detail the structure and functions of the J-TEXT CCS system as well as the performance in practical implementation. Results obtained from both commissioning and routine operations show that the J-TEXT CCS system can offer a satisfactory and effective control that is reliable and stable. The J-TEXT tokamak achieved high-quality performance in its first-ever experimental campaign with this CCS system.

Flux surface shaping effects on tokamak edge turbulence and flows

Energy Technology Data Exchange (ETDEWEB)

Kendl, A. [Innsbruck Univ., Institut fuer Theoretische Physik, Association EURATOM (Austria); Scott, B.D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching bei Muenchen (Germany)

2004-07-01

The influence of shaping of magnetic flux surfaces in tokamaks on gyro-fluid edge turbulence is studied numerically. Magnetic field shaping in tokamaks is mainly due to elongation, triangularity, shift and the presence of a divertor X-point. A series of tokamak configurations with varying elongation 1 {<=} {kappa} {>=} 2 and triangularity 0 {<=} {delta} {<=} 0.4, and an actual ASDEX Upgrade divertor configuration are obtained with the equilibrium code HELENA and implemented into the gyro-fluid turbulence code GEM. The study finds minimal impact on the zonal flow physics itself, but strong impact on the turbulence and transport. (authors)

Flux surface shaping effects on tokamak edge turbulence and flows

International Nuclear Information System (INIS)

Kendl, A.; Scott, B.D.

2004-01-01

The influence of shaping of magnetic flux surfaces in tokamaks on gyro-fluid edge turbulence is studied numerically. Magnetic field shaping in tokamaks is mainly due to elongation, triangularity, shift and the presence of a divertor X-point. A series of tokamak configurations with varying elongation 1 ≤ κ ≥ 2 and triangularity 0 ≤ δ ≤ 0.4, and an actual ASDEX Upgrade divertor configuration are obtained with the equilibrium code HELENA and implemented into the gyro-fluid turbulence code GEM. The study finds minimal impact on the zonal flow physics itself, but strong impact on the turbulence and transport. (authors)

Boundary Plasma Turbulence Simulations for Tokamaks

International Nuclear Information System (INIS)

Xu, X.; Umansky, M.; Dudson, B.; Snyder, P.

2008-05-01

The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T e ; T i ) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics

Global gyrokinetic simulation of tokamak transport

International Nuclear Information System (INIS)

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

1998-10-01

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

ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

International Nuclear Information System (INIS)

Berry, Jan; Ferrada, Juan J.; Curd, Warren; Dell Orco, Giovanni; Barabash, Vladimir; Kim, Seokho H.

2011-01-01

During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support

Tore Supra. Basic design Tokamak system

International Nuclear Information System (INIS)

Aymar, R.; Bareyt, B.; Bon Mardion, G.

1980-10-01

This document describes the basic design for the main components of the Tokamak system of Tora Supra. As such, it focuses on the engineering problems, and refers to last year report on Tora Supra (EUR-CEA-1021) for objectives and experimental programme of the apparatus on one hand, and for qualifying tests of the main technical solutions on the other hand. Superconducting toroidal field coil system, vacuum vessels and radiation shields, poloidal field system and cryogenic system are described

Equilibrium Reconstruction in EAST Tokamak

International Nuclear Information System (INIS)

Qian Jinping; Wan Baonian; Shen Biao; Sun Youwen; Liu Dongmei; Xiao Bingjia; Ren Qilong; Gong Xianzu; Li Jiangang; Lao, L. L.; Sabbagh, S. A.

2009-01-01

Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign. (magnetically confined plasma)

Central control system for the EAST tokamak

International Nuclear Information System (INIS)

Sun Xiaoyang; Ji Zhenshan; Wu Yicun; Luo Jiarong

2008-01-01

The architecture, the main function and the design scheme of the central control system and the collaboration system of EAST tokamak are described. The main functions of the central control system are to supply a union control interface for all the control, diagnoses, and data acquisition (DAQ) subsystem and it is also designed to synchronize all those subsystem. (authors)

Data processing system for spectroscopy at Novillo Tokamak; Sistema de procesamiento de datos para espectroscopia en el Tokamak Novillo

Energy Technology Data Exchange (ETDEWEB)

Ortega C, G.; Gaytan G, E. [Instituto Tecnologico de Toluca, Instituto nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1998-07-01

Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

ANTWKB: a code for the simulation of ion cyclotron antennas in tokamaks

International Nuclear Information System (INIS)

Brambilla, M.

1995-04-01

We have developed a code which evaluates the complex input impedance, the loading, and the spectral distribution of the launched power, of metallic antennas for ion cyclotron heating of large tokamak plasmas. The current distribution along the conductors is obtained selfconsistently from a variational method. The plasma response is evaluated assuming that the WKB approximation can be used already at the plasma edge, thereby avoiding the lengthy integration of the wave equations in the plasma. This makes possible systematic scans over frequency or other parameters, while retaining a sufficiently large number of Fourier components in the radiated fields to ensure convergence of both the resistive and reactive part of the power. Optionally, the code can evaluate the antenna response in vacuum or with a dummy load, for comparison with test bank measurements. We have applied the code to a few antennas of practical interest. The code reproduces accurately the expected transmission-line-like behaviour of a simple feeder-to-short antenna, and reasonably well the measured properties of the folded antenna of the ASDEX Upgrade ICRF experiment. This antenna is found to have particularly favourable properties, since its outer conductors present to the plasma a relatively uniform current over a broad range of frequencies, which, moreover, is always larger than in the return conductors. The loading of the ''violin antenna'' recently proposed for use in ITER is found to be satisfactory in the vicinity of antenna resonance, although rather poor at other frequencies. In the case of simple strap antennas replacing the short by an adjustable capacity, as in TORE SUPRA, is confirmed to be a good way of optimizing the loading. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

PPPL tokamak program

International Nuclear Information System (INIS)

Furth, H.P.

1984-10-01

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

Numerical studies of edge localized instabilities in tokamaks

International Nuclear Information System (INIS)

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

2002-01-01

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

Study of intelligent system for control of the tokamak-ETE plasma positioning

International Nuclear Information System (INIS)

Barbosa, Luis Filipe de Faria Pereira Wiltgen

2003-01-01

The development of an intelligent neural control system of the neural type, capable to perform real time control of the plasma displacement in the experiment tokamak spheric - ETE (spherical tokamak experiment ) is presented. The ETE machine is in operation since Nov 2000, in the LAP - Plasma Associated Laboratory of the Brazilian Institute on Spatial Research (INPE) in Sao Jose dos Campos, S P, Brazil. The experiment is dedicated to study the magnetic confinement of a fusion plasma in a configuration favorable for the construction of future reactors. Nuclear fusion constitutes a renewable energy source with low environmental impact, which uses atomic energy in pacific applications for the sustainable development of humanity. One of the important questions for the attainment of fusion relates to the stability of the plasma and control of its position during the reactor operation. Therefore, the development of systems to control the plasma in tokamaks constitutes a necessary technological advance for the feasibility of nuclear fusion. In particular, the research carried out in this thesis concerns the proposal of a system to control the vertical displacement of the plasma in the ETE tokamak, aiming to obtain steady pulses in this machine. A Magnetic Levitation system (Mag Lev) was developed as part of this work, allowing to study the nonlinear behavior of a device that, from the aspect of position control, is similar (analogous) to the plasma in the ETE tokamak, This magnetic levitation system was designed, mathematically modeled and built in order to test both classical and intelligent type controllers. The results of this comparison are very promising for the use of intelligent controllers in the ETE tokamak as well as other control applications. (author)

Finite element circuit theory of the numerical code EDDYMULT for solving eddy current problems in a multi-torus system

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Ozeki, Takahisa

1986-07-01

The finite element circuit theory is extended to the general eddy current problem in a multi-torus system, which consists of various torus conductors and axisymmetric coil systems. The numerical procedures are devised to avoid practical restrictions of computer storage and computing time, that is, the reduction technique of eddy current eigen modes to save storage and the introduction of shape function into the double area integral of mode coupling to save time. The numerical code EDDYMULT based on the theory is developed to use in designing tokamak device from the viewpoints of the evaluation of electromagnetic loading on the device components and the control analysis of tokamak equilibrium. (author)

Upgrade and benchmarking of the NIFS physics-engineering-cost code

International Nuclear Information System (INIS)

Dolan, T.J.; Yamazaki, K.

2004-07-01

The NIFS Physics-Engineering-Cost (PEC) code for helical and tokamak fusion reactors is upgraded by adding data from three blanket-shield designs, a new cost section based on the ARIES cost schedule, more recent unit costs, and improved algorithms for various computations. The PEC code is also benchmarked by modeling the ARIES-AT (advanced technology) tokamak and the ARIES-SPPS (stellarator power plant system). The PEC code succeeds in predicting many of the pertinent plasma parameters and reactor component masses within about 10%. There are cost differences greater than 10% for some fusion power core components, which may be attributed to differences of unit costs used by the codes. The COEs estimated by the PEC code differ from the COEs of the ARIES-AT and ARIES-SPPS studies by 5%. (author)

Axisymmetric instability in a noncircular tokamak

International Nuclear Information System (INIS)

Lipschultz, B.

1979-10-01

The stability of dee, inverse-dee and square cross section plasmas to axisymmetric modes has been investigated experimentally in Tokapole II, a tokamak with a four-null poloidal divertor. Experimental results are closely compared with predictions of two numerical stability codes - the PEST code (ideal MHD, linear stability) adapted to tokapole geometry and a code which follows the nonlinear evolution of shapes similar to tokapole equilibria

DIII-D tokamak control and neutral beam computer system upgrades

International Nuclear Information System (INIS)

Penaflor, B.G.; McHarg, B.B.; Piglowski, D.A.; Pham, D.; Phillips, J.C.

2004-01-01

This paper covers recent computer system upgrades made to the DIII-D tokamak control and neutral beam computer systems. The systems responsible for monitoring and controlling the DIII-D tokamak and injecting neutral beam power have recently come online with new computing hardware and software. The new hardware and software have provided a number of significant improvements over the previous Modcomp AEG VME and accessware based systems. These improvements include the incorporation of faster, less expensive, and more readily available computing hardware which have provided performance increases of up to a factor 20 over the prior systems. A more modern graphical user interface with advanced plotting capabilities has improved feedback to users on the operating status of the tokamak and neutral beam systems. The elimination of aging and non supportable hardware and software has increased overall maintainability. The distinguishing characteristics of the new system include: (1) a PC based computer platform running the Redhat version of the Linux operating system; (2) a custom PCI CAMAC software driver developed by general atomics for the kinetic systems 2115 serial highway card; and (3) a custom developed supervisory control and data acquisition (SCADA) software package based on Kylix, an inexpensive interactive development environment (IDE) tool from borland corporation. This paper provides specific details of the upgraded computer systems

An advanced plasma control system for the DIII-D tokamak

International Nuclear Information System (INIS)

Ferron, J.R.; Kellman, A.; McKee, E.; Osborne, T.; Petrach, P.; Taylor, T.S.; Wight, J.; Lazarus, E.

1991-11-01

An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Development of this system is expected to lead to control system technology appropriate for use on future tokamaks such as ITER and BPX. The digital system will allow for increased precision in shape control through real time adjustment of the control algorithm to changes in the shape and discharge parameters such as β p , ell i and scrape-off layer current. The system will be used for research on real time optimization of discharge performance for disruption avoidance, current and pressure profile control, optimization of rf antenna loading, or feedback on heat deposition patterns through divertor strike point position control, for example. Shape control with this system is based on linearization near a target shape of the controlled parameters as a function of the magnetic diagnostic signals. This digital system is unique in that it is designed to have the speed necessary to control the unstable vertical motion of highly elongated tokamak discharges such as those produced in DIII-D and planned for BPX and ITER. a 40 MHz Intel i860 processor is interfaced to up to 112 channels of analog input signals. The commands to the poloidal field coils can be updated at 80 μs intervals for the control of vertical position with a delay between sampling of the analog signal and update of the command of less than 80 μs

Modular pulse sequencing in a tokamak system

International Nuclear Information System (INIS)

Chew, A.C.; Lee, S.; Saw, S.H.

1992-01-01

Pulse technique applied in the timing and sequencing of the various part of the MUT tokamak system are discussed. The modular architecture of the pulse generating device highlights the versatile application of the simple physical concepts in precise and complicated research experiment. (author)

Design and construction of Alborz tokamak vacuum vessel system

International Nuclear Information System (INIS)

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

2012-01-01

Highlights: ► The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. ► As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. ► A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. ► Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

Dynamic simulations of the cryogenic system of a tokamak

International Nuclear Information System (INIS)

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

2015-01-01

In a tokamak plasma confinement is achieved through high magnetic fields generated by superconductive coils that need to be cooled down to 4.4 K with a forced flow of supercritical Helium. Tokamak's coil system works cyclically and so it is subject to pulsed heat loads which have to be handled by the refrigerator. This latter has to be sized on the average power value and not according to the peak to limit investment and operation costs and hence the heat load needs to be smoothed. CEA Grenoble is in charge of providing the cryogenic system for the Japanese tokamak JT60-SA, currently under construction in Naka (Japan). Hence, in order to model and study the smoothing strategies, an experimental set up: HELIOS (Helium Loop for high load smoothing) has been built. This is a scaled down model (1:20) of the helium distribution system whose main components are a saturated helium bath and a supercritical helium loop. This large installation can reproduce conditions of pressure, temperature and transport times, similar to those expected in the cooling circuits of the central solenoid superconducting magnets of JT-60SA. The peak loads representative of the tokamak operation have been reproduced and smoothed before they arrive in the refrigerator, by means of a saturated helium bath (thermal reservoir). A dynamic modelling of the cryogenic system is presented, with results on the pulsed load scenarios. All the simulations have been performed with EcosimPro software developed and the cryogenic library: CRYOLIB. This document is made up of an abstract and the slides of the presentation

Simulation of MHD instability effects on burning plasma transport with ITB in tokamak and helical reactors

International Nuclear Information System (INIS)

Yamazaki, K.; Yamada, I.; Taniguchi, S.; Oishi, T.

2009-01-01

Full text: The high performance plasma behavior is required to realize economic and environmental-friendly fusion reactors compatible with conventional power plant systems. To improve plasma confinement, the formation of internal transport barrier (ITB) is anticipated, and its behavior is analyzed by the simulation code TOTAL (Toroidal Transport Linkage Analysis). This TOTAL code comprises a 2- or 3-dimensional equilibrium and 1-dimensional predictive transport code for both tokamak and helical systems. In the tokamak code TOTAL-T, the external current drive, bootstrap current, sawtooth oscillation, ballooning mode and neoclassical tearing mode (NTM) analyses are included. The steady-state burning plasma operation is achieved by the feedback control of pellet injection fuelling and external heating power control. The impurity dynamics of iron and tungsten is also included in this code. The NTM effects are evaluated using the modified Rutherford Model with the stabilization of the ECCD current drive. The excitation of m=2/n=1 NTM leads to the 20 % reduction in the central temperature in ITER-like reactors. Recently, the external non-resonant helical field application is analyzed and its stabilization properties are evaluated. The pellet injection effects on ITB formation is also clarified in tokamak and helical plasmas. Relationship between sawtooth oscillation and impurity ejection is recently simulated in comparison with experimental data. In this conference, we will show above-stated new results on MHD instability effects on burning plasma transport. (author)

Software development of the KSTAR Tokamak Monitoring System

International Nuclear Information System (INIS)

Kim, K.H.; Lee, T.G.; Baek, S.; Lee, S.I.; Chu, Y.; Kim, Y.O.; Kim, J.S.; Park, M.K.; Oh, Y.K.

2008-01-01

The Korea Superconducting Tokamak Advanced Research (KSTAR) project, which is constructing a superconducting Tokamak, was launched in 1996. Much progress in instrumentation and control has been made since then and the construction phase will be finished in August 2007. The Tokamak Monitoring System (TMS) measures the temperatures of the superconducting magnets, bus-lines, and structures and hence monitors the superconducting conditions during the operation of the KSTAR Tokamak. The TMS also measures the strains and displacements on the structures in order to monitor the mechanical safety. There are around 400 temperature sensors, more than 240 strain gauges, 10 displacement gauges and 10 Hall sensors. The TMS utilizes Cernox sensors for low temperature measurement and each sensor has its own characteristic curve. In addition, the TMS needs to perform complex arithmetic operations to convert the measurements into temperatures for each Cernox sensor for this large number of monitoring channels. A special software development effort was required to reduce the temperature conversion time and multi-threading to achieve the higher performance needed to handle the large number of channels. We have developed the TMS with PXI hardware and with EPICS software. We will describe the details of the implementations in this paper

Ion cyclotron system design for KSTAR tokamak

International Nuclear Information System (INIS)

Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H.

1998-05-01

The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs

Ion cyclotron system design for KSTAR tokamak

Energy Technology Data Exchange (ETDEWEB)

Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H

1998-05-01

The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs.

ZORNOC: a 1 1/2-D tokamak data analysis code for studying noncircular high beta plasmas

International Nuclear Information System (INIS)

Zurro, B.; Wieland, R.M.; Murakami, M.; Swain, D.W.

1980-03-01

A new tokamak data analysis code, ZORNOC, was developed to study noncircular, high beta plasmas in the Impurity Study Experiment (ISX-B). These plasmas exhibit significant flux surface shifts and elongation in both ohmically heated and beam-heated discharges. The MHD equilibrium flux surface geometry is determined by solving the Grad-Shafranov equation based on: (1) the shape of the outermost flux surface, deduced from the magnetic loop probes; (2) a pressure profile, deduced by means of Thomson scattering data (electrons), charge exchange data (ions), and a Fokker-Planck model (fast ions); and (3) a safety factor profile, determined from the experimental data using a simple model (Z/sub eff/ = const) that is self-consistently altered while the plasma equilibrium is iterated. For beam-heated discharches the beam deposition profile is determined by means of a Monte Carlo scheme and the slowing down of the fast ions by means of an analytical solution of the Fokker-Planck equation. The code also carries out an electron power balance and calculates various confinement parameters. The code is described and examples of its operation are given

Economic evaluation of fissile fuel production using resistive magnet tokamaks

International Nuclear Information System (INIS)

Doyle, J.C. Jr.

1985-06-01

The application of resistive magnet tokamaks to fissile fuel production has been studied. Resistive magnets offer potential advantages over superconducting magnets in terms of robustness, less technology development required and possibility of demountable joints. Optimization studies within conservatively specified constraints for a compact machine result in a major radius of 3.81 m and 618 MW fusion power and a blanket space envelope of 0.35 m inboard and 0.75 m outboard. This machine is called the Resistive magnet Tokamak Fusion Breeder (RTFB). A computer code was developed to estimate the cost of the resistive magnet tokamak breeder. This code scales from STARFIRE values where appropriate and calculates costs of other systems directly. The estimated cost of the RTFB is $3.01 B in 1984 dollars. The cost of electricity on the same basis as STARFIRE is 42.4 mills/kWhre vs 44.9 mills/kWhre for STARFIRE (this does not include the fuel value or fuel cycle costs for the RTFB). The breakeven cost of U 3 O 8 is $150/lb when compared to a PWR on the once through uranium fuel cycle with no inflation and escalation. On the same basis, the breakeven cost for superconducting tokamak and tandem mirror fusion breeders is $160/lb and $175/lb. Thus, the RTFB appears to be competitive in breakeven U 3 O 8 cost with superconducting magnet fusion breeders and offers the potential advantages of resistive magnet technology

Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code

Directory of Open Access Journals (Sweden)

Yutthapong Pinanroj

2014-04-01

Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.

Lithium beam diagnostic system on the COMPASS tokamak

Energy Technology Data Exchange (ETDEWEB)

Anda, G.; Bencze, A. [Wigner – RCP, HAS, Budapest (Hungary); Berta, M., E-mail: bertam@sze.hu [Institute of Plasma Physics AS CR, Prague (Czech Republic); Széchenyi István University, Győr (Hungary); Dunai, D. [Wigner – RCP, HAS, Budapest (Hungary); Hacek, P. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Krbec, J. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Réfy, D.; Krizsanóczi, T.; Bató, S.; Ilkei, T.; Kiss, I.G.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, Budapest (Hungary)

2016-10-15

Highlights: • Li-beam diagnostic system on the COMPASS tokamak is an improved and compact system to allow testing of Atomic Beam Probe. • The possibility to measure background corrected density profiles on the few microseconds time scale. • First Li-beam diagnostic system with recirculating neutralizer. • The system includes the redesigned ion source with longer lifetime. - Abstract: An improved lithium beam based beam emission spectroscopy system – installed on COMPASS tokamak – is described. The beam energy enhanced up to 120 keV for Atomic Beam Probe measurement. The size of the ion source is doubled, using a newly developed thermionic heater instead of the conventionally used heating (tungsten or molybdenum) filament. The neutralizer is also improved. It produces the same sodium vapor in a cell but minimize the loss condensing the vapor on a cold surface which is led back (in fluid state) into the sodium oven. This way we call it recirculating neutralizer. The observation system consists of a CCD camera and an avalanche photodiode array.

Numerical simulation of edge plasma in tokamak

International Nuclear Information System (INIS)

Chen Yiping; Qiu Lijian

1996-02-01

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

A cryogenic system for TIBER II [Tokamak Ignition/Burn Experimental Reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.

1987-01-01

Phase II of the Tokamak Ignition/Burn Experimental Reactor (TIBER II) study describes one option for a small, economical, next-generation tokamak [1,2]. Because of its small size, minimum shielding is used between the plasma and the toroidal-field (TF) coils. Consequently, a large cryogenic system (approximately 70 kW at 4.5 K) capable of delivering forced-flow helium is required. This paper describes a cryogenic system that meets this requirement and includes TIBER-II requirements. 3 refs

Turbulence and abnormal transport in tokamak plasmas

International Nuclear Information System (INIS)

Garbet, X.

1988-06-01

The objective of this thesis is the study of plasma microinstabilities in linear and nonlinear tokamak regime. After a brief review of experimental results the theoretical tools used in this study are presented. A variational method founded on the existence of angular variables system and on action for charged particles in tokamak configurations is detailed. The correspondent functional extreme with regard to fluctuating electromagnetic field, is calculated analytically with taking into account the toroidal geometry. A numerical code, TORRID, has been constructed on this principle and the main instabilities, particularly ionic instabilities and microtearing, has been linearly studied. The most simple non linear methods are rewieved and applied at the microtearing instabilities. The quasilinear transport coefficients are deducted of an entropy minimum production principle. The ionic thermic conductivity and the viscosity are calculated for an ionic turbulence [fr

Analysis of EAST tokamak cryostat anti-seismic performance

International Nuclear Information System (INIS)

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

2014-01-01

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

The timing system on the J-TEXT tokamak

International Nuclear Information System (INIS)

Zheng, Wei; Zhang, Ming; Zhuang, Ge; Ding, Tonghai; Huang, Fuqiang; Shan, Lingjie

2014-01-01

Highlights: •The timing system achieved tree structured timing network with only one type of timing module. •This system is integrated into J-TEXT COADC which is an EPICS based control system. •This system handles multiple timing sequences and events. •This system has been deployed on J-TEXT and working properly in daily experiments. -- Abstract: This paper describes the timing system designed to control the operation time-sequence and to generate clocks for various sub-systems on J-TEXT tokamak. The J-TEXT timing system is organized as a distributed system which is connected by a tree-structured optical fiber network. It can generate delayed triggers and gate signals (0 μs–4000 s), while providing reference clocks for other sub-systems. Besides, it provides event handling and timestamping functions. It is integrated into the J-TEXT Control, Data Access and Communication (J-TEXT CODAC) system, and it can be monitored and configured by Experimental Physics and Industrial Control System (EPICS). The configuration of this system including tree-structured network is managed in XML files by dedicated management software. This system has already been deployed on J-TEXT tokamak and it is serving J-TEXT in daily experiments

Pneumatic hydrogen pellet injection system for the ISX tokamak

International Nuclear Information System (INIS)

Milora, S.L.; Foster, C.A.

1979-01-01

We describe the design and operation of the solid hydrogen pellet injection system used in plasma refueling experiments on the ISX tokamak. The gun-type injector operates on the principle of gas dynamic acceleration of cold pellets confined laterally in a tube. The device is cooled by flowing liquid helium refrigerant, and pellets are formed in situ. Room temperature helium gas at moderate pressure is used as the propellant. The prototype device injected single hydrogen pellets into the tokamak discharge at a nominal 330 m/s. The tokamak plasma fuel content was observed to increase by (0.5--1.2) x 10 19 particles subsequent to pellet injection. A simple modification to the existing design has extended the performance to 1000 m/s. At higher propellant operating pressures (28 bars), the muzzle velocity is 20% less than predicted by an idealized constant area expansion process

Implementing a finite-state off-normal and fault response system for disruption avoidance in tokamaks

Science.gov (United States)

Eidietis, N. W.; Choi, W.; Hahn, S. H.; Humphreys, D. A.; Sammuli, B. S.; Walker, M. L.

2018-05-01

A finite-state off-normal and fault response (ONFR) system is presented that provides the supervisory logic for comprehensive disruption avoidance and machine protection in tokamaks. Robust event handling is critical for ITER and future large tokamaks, where plasma parameters will necessarily approach stability limits and many systems will operate near their engineering limits. Events can be classified as off-normal plasmas events, e.g. neoclassical tearing modes or vertical displacements events, or faults, e.g. coil power supply failures. The ONFR system presented provides four critical features of a robust event handling system: sequential responses to cascading events, event recovery, simultaneous handling of multiple events and actuator prioritization. The finite-state logic is implemented in Matlab®/Stateflow® to allow rapid development and testing in an easily understood graphical format before automated export to the real-time plasma control system code. Experimental demonstrations of the ONFR algorithm on the DIII-D and KSTAR tokamaks are presented. In the most complex demonstration, the ONFR algorithm asynchronously applies ‘catch and subdue’ electron cyclotron current drive (ECCD) injection scheme to suppress a virulent 2/1 neoclassical tearing mode, subsequently shuts down ECCD for machine protection when the plasma becomes over-dense, and enables rotating 3D field entrainment of the ensuing locked mode to allow a safe rampdown, all in the same discharge without user intervention. When multiple ONFR states are active simultaneously and requesting the same actuator (e.g. neutral beam injection or gyrotrons), actuator prioritization is accomplished by sorting the pre-assigned priority values of each active ONFR state and giving complete control of the actuator to the state with highest priority. This early experience makes evident that additional research is required to develop an improved actuator sharing protocol, as well as a methodology to

Burnup calculation for a tokamak commercial hybrid reactor

International Nuclear Information System (INIS)

Feng Kaiming; Xie Zhongyou

1990-08-01

A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

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

Science.gov (United States)

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

2013-10-01

The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R security systems for MEDUSA-CR device. The interface with the control and data acquisition systems based on National Instruments (NI) software (LabView) and hardware (on loan to our laboratory via NI-Costa Rica) are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

Resistive MHD studies of high-β-tokamak plasmas

International Nuclear Information System (INIS)

Lynch, V.E.; Carreras, B.A.; Hicks, H.R.; Holmes, J.A.; Garcia, L.

1981-01-01

Numerical calculations have been performed to study the MHD activity in high-β tokamaks such as ISX-B. These initial value calculations built on earlier low β techniques, but the β effects create several new numerical issues. These issues are discussed and resolved. In addition to time-stepping modules, our system of computer codes includes equilibrium solvers (used to provide an initial condition) and output modules, such as a magnetic field line follower and an X-ray diagnostic code. The transition from current driven modes at low β to predominantly pressure driven modes at high β is described. The nonlinear studies yield X-ray emissivity plots which are compared with experiment

Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor

International Nuclear Information System (INIS)

Kim, Seokho H.; Berry, Jan

2011-01-01

U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclear pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.

Modular coils and finite-β operation of a quasi-axially symmetric tokamak

International Nuclear Information System (INIS)

Drevlak, M.

1998-01-01

Quasi-axially symmetric tokamaks (QA tokamaks) are an extension of the conventional tokamak concept. In these devices the magnetic field strength is independent of the generalized toroidal magnetic co-ordinate even though the cross-sectional shape changes. An optimized plasma equilibrium belonging to the class of QA tokamaks has been proposed by Nuehrenberg. It features the small aspect ratio of a tokamak while allowing part of the rotational transform to be generated by the external field. In this article, two particular aspects of the viability of QA tokamaks are explored, namely the feasibility of modular coils and the possibility of maintaining quasi-axial symmetry in the free-boundary equilibria obtained with the coils found. A set of easily feasible modular coils for the configuration is presented. It was designed using the extended version of the NESCOIL code (MERKEL, P., Nucl. Fusion 27 (1987) 867). Using this coil system, free-boundary calculations of the plasma equilibrium were carried out using the NEMEC code (HIRSHMAN, S.P., VAN RIJ, W.I., MERKEL, P., Comput. Phys. Commun. 43 (1986) 143). It is observed that the effects of finite β and net toroidal plasma current can be compensated for with good precision by applying a vertical magnetic field and by separately adjusting the currents of the modular coils. A set of fully three dimensional (3-D) auxiliary coils is proposed to exert control on the rotational transform in the plasma. Deterioration of the quasi-axial symmetry induced by the auxiliary coils can be avoided by adequate adjustment of the currents in the primary coils. Finally, the neoclassical transport properties of the configuration are examined. It is observed that optimization with respect to confinement of the alpha particles can be maintained at operation with finite toroidal current if the aforementioned corrective measures are used. In this case, the neoclassical behaviour is shown to be very similar to that of a conventional tokamak

Study of a compact reversed shear Tokamak reactor

International Nuclear Information System (INIS)

Okano, K.; Asaoka, Y.; Tomabechi, K.; Yoshida, T.; Hiwatari, R.; Ogawa, Y.; Tokimatsu, K.; Yamamoto, T.; Inoue, N.; Murakami, Y.

1998-01-01

A reversed shear configuration, which was observed recently in some tokamak experiments, might have a possibility to realize compact and cost-competitive tokamak reactors. In this study, a compact (low cost) commercial reactor based on the shear reversed high beta equilibrium with β N =5.5, is considered, namely the compact reversed shear tokamak, CREST-1. The CREST-1 is designed with a moderate aspect ratio (R/a=3.4), which will allow us to experimentally develop this CREST concept by ITER. This will be very advantageous with regard to the fusion development strategy. The current profile for the reversed shear operation is sustained and controlled in steady state by bootstrap (88%), beam and r driven currents, which are calculated by a neo-classical model code in 3D geometry. The MHD stability has been checked by an ideal MHD stability analysis code (ERATO) and it has been confirmed that the ideal low n kink, ballooning and Mercier modes are stable while a closed conductive shell is required for stability. Such a compact tokamak can be cost-competitive as an electric power source in the 21st century and it is one possible scenario in realizing a commercial fusion reactor beyond the ITER project. (orig.)

Digital controlled pulsed electric system of the ETE tokamak. First report

International Nuclear Information System (INIS)

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

1997-01-01

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

Tokamak building-design considerations for a large tokamak device

International Nuclear Information System (INIS)

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

1981-01-01

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

Distributed digital real-time control system for the TCV tokamak and its applications

NARCIS (Netherlands)

Anand, H.; Galperti, C.; Coda, S.; Duval, B.P.; Felici, F.; Blanken, T.; Maljaars, E.; Moret, J.M.; Sauter, O.; Goodman, T.P.; Kim, D.

2017-01-01

A key feature of the new digital plasma control system installed on the TCV (Tokamak à Configuration Variable) tokamak is its possibility to rapidly design, test and deploy real-time algorithms. It accommodates hundreds of diagnostic inputs and actuator outputs, and offers the possibility to design

Calculations of axisymmetric stability of tokamak plasmas with active and passive feedback

International Nuclear Information System (INIS)

Ward, D.J.; Jardin, S.C.; Cheng, C.Z.

1991-07-01

A new linear MHD stability code, NOVA-W, has been developed in order to study feedback stabilization of the axisymmetric mode in deformable tokamak plasmas. The NOVA-W code is a modification of the non-variational MHD stability code NOVA that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The passive stability predictions of the code have been tested both against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. Active feedback calculations are performed for the CIT tokamak design demonstrating the effect of varying the position of the flux loops that provide the measurements of vertical displacement. The results compare well with those computed earlier using a less efficient nonlinear code. 37 refs., 13 figs

The Discharge Design of HL-2M with the Tokamak Simulation Code (TSC)

International Nuclear Information System (INIS)

Yudong Pan; Jardin, S.C.; Kes, C.

2007-01-01

We present results on the discharge design of the HL-2M tokamak, which is to be an upgrade to the existing HL-2A tokamak. We present simulation results for complete 5-sec. discharges, both double null and lower single null, for both ohmic and auxiliary heated discharges. We also discuss the vertical stability properties of the device

Parametric requirements for noncircular tokamak commercial fusion plants

International Nuclear Information System (INIS)

Bourque, R.F.

1978-05-01

Systems analyses have been performed in order to determine the parameter ranges for economical operation of equilibrium commercial fusion power plants using noncircular tokamak reactors. The fully-integrated systems code SCOPE was employed in which costs of electricity were used as the critical dependent variable by which the effects of changes in system parameters were observed. The results of these studies show that many of the operating parameters required for economical operation can be considerably relaxed from what has been thought necessary. For example: (1) Neutron wall loadings over about 2 megawatts per square meter are unnecessary for economical tokamak operation; (2) electrical power levels beyond 1000 MW(e) are not required. In fact, power levels as low as 100 MW(e) may be acceptable; (3) total beta values of 10 to 12 percent are adequate for economic operation; (4) duty cycles as low as 50 percent have only a small effect on plant economics; (5) an acceptable first wall lifetime is 15 MW-yr per square meter

Research tokamak system with multi-mode discharges using inverter power supply

International Nuclear Information System (INIS)

Kojima, Hiroki; Kobayashi, Masahiro; Takagi, Makoto; Takamura, Shuichi; Tashiro, Kenji

1999-01-01

In Current Sustaining Tokamak in Nagoya university (CSTN)-IV research tokamak system using a compact 40kHz pulse width modulation (PWM) inverter power supply, which is controlled through LabVIEW program, we construct a new tokamak discharge system with multi-mode including a stable alternating current discharge and a high-repetition high-duty one. These discharge modes can be operated continuously for as long as 60sec. The continuous discharge with long duration is able to simulate the important physical and chemical processes of long time discharges in fusion devices, in which the heat load to the wall and the particle balance in the plasma-wall system are crucial topics in order to realize a long pulse fusion reactor, like ITER. Employing ergodic divertor (ED) is one of tools to control the particle balance and the heat load to the wall. In addition, we installed another inverter power supply to generate a rotating magnetic perturbation for dynamic ergodic divertor (DED) with the appropriate measurement system so that we may carry out experiments on heat and particle control with DED at long time operation. (author)

Tokamak experiments

International Nuclear Information System (INIS)

Robinson, D.C.

1987-01-01

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

Real-time software for the COMPASS tokamak plasma control

International Nuclear Information System (INIS)

Valcarcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Sartori, F.; Janky, F.; Cahyna, P.; Hron, M.; Panek, R.

2010-01-01

The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 μs. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.

Real-time software for the COMPASS tokamak plasma control

Energy Technology Data Exchange (ETDEWEB)

Valcarcel, D.F., E-mail: danielv@ipfn.ist.utl.p [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic)

2010-07-15

The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 {mu}s. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.

Impact of maximum TF magnetic field on performance and cost of an advanced physics tokamak

International Nuclear Information System (INIS)

Reid, R.L.

1983-01-01

Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of variation in the maximum value of the field at the toroidal field (TF) coils on the performance and cost of a low q/sub psi/, quasi-steady-state tokamak. Marginal ignition, inductive current startup plus 100 s of inductive burn, and a constant value of epsilon (inverse aspect ratio) times beta poloidal were global conditions imposed on this study. A maximum TF field of approximately 10 T was found to be appropriate for this device

ICRF [Ion Cyclotron Range of Frequencies] heating and antenna coupling in a high beta tokamak

International Nuclear Information System (INIS)

Elet, R.S.

1988-01-01

Maxwell's Equations are solved in two-dimensions for the electromagnetic fields in a toroidal cavity using the cold plasma fluid dielectric tensor in the Ion Cyclotron Range of Frequencies (ICRF). The Vector Wave Equation is transformed to a set of two, coupled second-order partial differential equations with inhomogeneous forcing functions which model a wave launcher. The resulting equations are finite differenced and solved numerically with a complex banded matrix algorithm on a Cray-2 computer using a code described in this report. This code is used to study power coupling characteristics of a wave launcher for low and high beta tokamaks. The low and high beta equilibrium tokamak magnetic fields applied in this model are determined from analytic solutions to the Grad-Shafranov equation. The code shows good correspondence with the results of low field side ICRF heating experiments performed on the Tokamak of Fontenay-Aux-Roses (TFR). Low field side and high field side antenna coupling properties for ICRF heating in the Columbia High Beta Tokamak (HBT) experiment are calculated with this code. Variations of antenna position in the tokamak, ionic concentration and plasma density, and volume-averaged beta have been analyzed for HBT. It is found that the location of the antenna with respect to the plasma has the dominant role in the design of an ICRF heating experiment in HBT. 10 refs., 52 figs., 13 tabs

Characteristics of ion Bernstein wave heating in JIPPT-II-U tokamak

International Nuclear Information System (INIS)

Okamoto, M.; Ono, M.

1985-11-01

Using a transport code combined with an ion Bernstein wave tokamak ray tracing code, a modelling code for the ion Bernstein wave heating has been developed. Using this code, the ion Bernstein wave heating experiment on the JIPPT-II-U tokamak has been analyzed. It is assumed that the resonance layer is formed by the third harmonic of deuterium-like ions, such as fully ionized carbon, and oxygen ions near the plasma center. For wave absorption mechanisms, electron Landau damping, ion cyclotron harmonic damping, and collisional damping are considered. The characteristics of the ion Bernstein wave heating experiment, such as the ion temperature increase, the strong dependence of the quality factor on the magnetic field strength, and the dependence of the ion temperature increment on the input power, are well reproduced

Preliminary Design of Alborz Tokamak

Science.gov (United States)

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

2012-04-01

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

Application of SNAM to the nuclear analysis of EAST Tokamak

International Nuclear Information System (INIS)

Hu, H.; Chen, M.; Zeng, Q.; Wu, Y.

2007-01-01

EAST (Experimental Advanced Superconducting Tokamak) is the first non-round cross section complete superconducting fusion experimental tokamak device built at the Institute of Plasma Physics, Chinese Academy of Sciences. Since 2.45MeV neutrons from D-D fusion reaction and 14.1MeV neutrons from D-T fusion reaction can both be generated during the DD plasma discharge, the distribution of neutron flux and nuclear heat in the device has an important effect on the nuclear and safety analysis. For radiation transport calculations, the main calculation tool is Monte Carlo transport code (MCNP) which has been used to give specific nuclear responses in complex geometries. However, the discrete ordinate transport code (SN code) is more effective to calculate the distribution of neutron flux and nuclear heat in the whole device. It is a time-consuming and error prone task to prepare the neutronics model for SN code in manual way. And because of the self-limitation of SN method, most of SN codes only support some specified or regular geometries, such as cylindrical, Cartesian and tetrahedral geometry, etc. In practice, most of models are composed of irregular solids and can not be supported by SN code. A more efficient solution is to shift the geometric modeling into a computer aided design (CAD) system and use an interface for SN code to convert CAD model into the input file automatically. SNAM (SN Code Automatic Modeling) is an integrated interface code between CAD system and SN code. The CAD model can be automatically converted into the input file of SN code with SNAM. On the contrary, SNAM can convert already existed input file of SN code into CAD model, which can be used to check, analyze, modify and reuse the model for the user. In this contribution, the process of converting EAST CAD model to the input file of SN code with SNAM is described. The distribution of neutron flux and nuclear heat in the whole device are calculated using the input file, and the results of the

Magnetic field structure near the plasma boundary in helical systems and divertor tokamaks

International Nuclear Information System (INIS)

Nagasaki, Kazunobu; Itoh, Kimitaka

1990-02-01

Magnetic field structure of the scrape off layer (SOL) region in both helical systems and divertor tokamaks is studied numerically by using model fields. The connection length of the field line to the wall is calculated. In helical systems, the connection length, L, has a logarithmic dependence on the distance from the outermost magnetic surface or that from the residual magnetic islands. The effect of axisymmetric fields on the field structure is also determined. In divertor tokamaks, the connection length also has logarithmic properties near the separatrix. Even when the perturbations, which resonate to rational surfaces near the plasma boundary, are added, logarithmic properties still remain. We compare the connection length of torsatron/helical-heliotron systems with that of divertor tokamaks. It is found that the former is shorter than the latter by one order magnitude with similar aspect ratio. (author)

D-D tokamak reactor assessment

International Nuclear Information System (INIS)

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

1983-01-01

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

Electrical probe diagnostic with fast data acquisition systems of the Novillo tokamak

International Nuclear Information System (INIS)

Lopez-Callejas, R.; Benitez-Read, J.S.; Longoria-Gandara, L.C.; Pacheco-Sotelo, J.O.; Valencia-Alvarado, R.; Tamayo, F.J.; Valdes, A.; Fernandez, M.C.; Serrano, F.

2001-01-01

This paper describes the electrical probe diagnostic system used to measure the relevant parameters of the plasma column generated in the Novillo tokamak, and an inexpensive and fast data acquisition system that consists of 16 independent channels, 8 bit resolution, sampling frequency of 500 kHz, and 8 kword memory per channel. A distributed, modular and transparent approach was used for designing the software-operated data acquisition, diagnostic, and capacitor banks triggering systems, satisfying the pulsed nature of the tokamak discharge. Through a personal computer, the experimental data are available to engineers and physicists in a centralized database

New DIII-D tokamak plasma control system

International Nuclear Information System (INIS)

Campbell, G.L.; Ferron, J.R.; McKee, E.; Nerem, A.; Smith, T.; Greenfield, C.M.; Pinsker, R.I.; Lazarus, E.A.

1992-09-01

A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and diverter power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the US Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter

A discussion on practicable scheme for machine control system of HL-2A tokamak

International Nuclear Information System (INIS)

Li Qiang; Song Xianming; Jiang Chao

2001-01-01

This is the modified version of The Preliminary Design on the Hardware and Nets of HL-2A Tokamak. In this version, centralized control as well as the field bus communication on HL-2A Tokamak is used. The hardware components and the operational theory are introduced. The questions of practice program, extensibility, centralized control, cooperation with other subsystems and the continuous adaptation of present device are all discussed. The budget of the system is detailed. To keep the step with the overall engineering constructions of HL-2A, suggestions of the time program are presented for the system design, instrument purchases, installation, construction, user program development and the final operation processes for the machine control system of HL-2A Tokamak

Multi-channel bolometer system on JFT-2M tokamak

International Nuclear Information System (INIS)

Tamai, Hiroshi; Maeno, Masaki; Matsuda, Toshiaki; Matoba, Tohru

1988-07-01

Multi-channel bolometer system is designed and installed to observe the radiation profile on JFT-2M tokamak. Sensor head is made of Thinistor, which is a kind of semiconductor, because it has the advantage of higher sensitivity of about one order of magnitude than the conventional metal foil bolometer and is suitable for the profile measurement in which the signal from the plasma is relatively small. The response and cooling characteristics of the bolometer sensor are suitable for the condition of JFT-2M tokamak plasma. Low noise circuit of bridge and differentiator is developed to optimize the signal to noise ratio in the JFT-2M operating condition. With use of the bolometer system, the radiation profile in joule heating plasma as well as additional heating plasma especially in H-mode plasma is successfully observed. (author)

Reconstruction of Magnetic Field Surfaces of the NOVILLO Tokamak by means of the 3D-MAPTOR Code

International Nuclear Information System (INIS)

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

2008-01-01

A 3-D code has been developed in order to simulate the magnetic field lines in circular cross-section tokamaks. The toroidal magnetic field can be obtained from the individual fields of circular coils arranged around the torus, or alternatively, as a ripple-less field, as well as the vertical field coils, and divertor-like coils. The poloidal field is provided by a given toroidal current density profile. Proposing initial conditions for a magnetic filed line, it is integrated along the toroidal angle coordinate, and the Poincare maps can be obtained at any desired cross section plane along the torus. Following this procedure, the code allows to explore the necessary current values for the existence of magnetic field surfaces, allowing for deviations from axial symmetry, such as ripple effects. Therefore it is a good design instrument, in which different parameters and arrangements of coils can be tested. On the other hand, the current signals from experimental devices can be used in order to reconstruct the behaviour of the magnetic field surfaces, including the q(r) profiles. The reconstruction properties of the code are shown in this work

A flexible software design to determine the plasma boundary in Damavand tokamak

International Nuclear Information System (INIS)

Ghadiri, R.; Esteki, M.H.; Sadeghi, Y.

2014-01-01

A plasma boundary reconstruction code has been designed by using current filament method to calculate the magnetic flux and consequently plasma boundary in Damavand tokamak. Hence, a computer-based code “The Plasma Boundary Reconstruction Code in Tokamak (PBRCT)” was developed to make a graphical user interface and to speed up the plasma boundary estimation algorithm. All required tools as the plasma boundary and magnetic surface display (MSD), error display, primary conditions and modeling panel as well as a search motor to determine a good position and number of the current filaments to find a precise model have been considered. The core is a 3000 lines Matlab code and the graphical user interface is 10,000 lines in C language. (author)

3D plasma fluid simulations in divertor tokamaks. Final technical report, 1993--1995

International Nuclear Information System (INIS)

Strauss, H.R.

1995-08-01

The main accomplishment of this grant was the development of a finite element time dependent magnetofluid code, FEMHD. The code is nonlinear and three dimensional. In the poloidal plane, the elemental cells of the mesh are triangles, which offer both simplicity and adaptability. In the third, toroidal, direction, there is an option of a standard staggered finite difference mesh, or Fourier transforms. The FEMHD code runs on several platforms, including Crays, UNIX workstations, and a parallel version runs on an IBM SP1. Several problems have been considered with the unstructured mesh FEMHD code. They are (1) MHD simulations in divertor tokamaks; (2) simulations of ELM-like ballooning modes in divertor tokamaks; and (3) reconnection and singular MHD equilibria

Catalogue of nuclear fusion codes - 1976

International Nuclear Information System (INIS)

1976-10-01

A catalogue is presented of the computer codes in nuclear fusion research developed by JAERI, Division of Thermonuclear Fusion Research and Division of Large Tokamak Development in particular. It contains a total of about 100 codes under the categories: Atomic Process, Data Handling, Experimental Data Processing, Engineering, Input and Output, Special Languages and Their Application, Mathematical Programming, Miscellaneous, Numerical Analysis, Nuclear Physics, Plasma Physics and Fusion Research, Plasma Simulation and Numerical Technique, Reactor Design, Solid State Physics, Statistics, and System Program. (auth.)

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-03-01

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

A synchronization system to digitalize TJ-1 Tokamak data

International Nuclear Information System (INIS)

Guasp, J.; Perez-Navarro, A.; Pacios, L.

1983-01-01

At TJ-1 Tokamak signals are stored on a 60-channel magnetic memory. In this report, a system to address those channels and synchronize readout is presented. Digitalized signals are stored in structured files on PDP-11/34 magnetic disks. (author)

Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

International Nuclear Information System (INIS)

Sager, G.T.; Wong, C.P.C.; Kapich, D.D.; McDonald, C.F.; Schleicher, R.W.

1993-11-01

The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankie and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed

Validation of equilibrium tools on the COMPASS tokamak

Energy Technology Data Exchange (ETDEWEB)

Urban, J., E-mail: urban@ipp.cas.cz [Institute of Plasma Physics ASCR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Appel, L.C. [CCFE, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); Artaud, J.F. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Faugeras, B. [Laboratoire J.A. Dieudonné, UMR 7351, Université de Nice Sophia-Antipolis, Parc Valrose, 06108 Nice Cedex 02 (France); Havlicek, J. [Institute of Plasma Physics ASCR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic); Komm, M. [Institute of Plasma Physics ASCR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Lupelli, I. [CCFE, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); Peterka, M. [Institute of Plasma Physics ASCR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic)

2015-10-15

Highlights: • Three equilibrium codes—EFIT++, FREEBIE and VacTH—have been successfully set up and validated on COMPASS. • FREEBIE can predictively calculate the equilibrium and corresponding poloidal field coil currents. • EFIT++ can reconstruct equilibria generated by FREEBIE from synthetic, optionally noisy diagnostic data. • VacTH is a promising tool for real time plasma shape reconstruction. • Optimized parameters are estimated for EFIT++ and VacTH by a statistical analysis. - Abstract: Various MHD (magnetohydrodynamic) equilibrium tools, some of which being recently developed or considerably updated, are used on the COMPASS tokamak at IPP Prague. MHD equilibrium is a fundamental property of the tokamak plasma, whose knowledge is required for many diagnostics and modelling tools. Proper benchmarking and validation of equilibrium tools is thus key for interpreting and planning tokamak experiments. We present here benchmarks and comparisons to experimental data of the EFIT++ reconstruction code (Appel et al., 2006), the free-boundary equilibrium code FREEBIE (Artaud and Kim, 2012), and a rapid plasma boundary reconstruction code VacTH (Faugeras et al., 2014). We demonstrate that FREEBIE can calculate the equilibrium and corresponding poloidal field (PF) coils currents consistently with EFIT++ reconstructions from experimental data. Both EFIT++ and VacTH can reconstruct equilibria generated by FREEBIE from synthetic, optionally noisy diagnostic data. Hence, VacTH is suitable for real-time control. Optimum reconstruction parameters are estimated.

Non-axisymmetric simulation of the vertical displacement event in tokamaks

International Nuclear Information System (INIS)

Lim, Y.Y.; Lee, J.K.; Shin, K.J.; Hur, M.S.

1999-01-01

Tokamak plasmas with highly elongated cross sections are subject to a vertical displacement event (VDE). The nonlinear magnetohydrodynamic (MHD) evolutions of tokamak plasmas during the VDE are simulated by a three-dimensional MHD code as a combination of N=0 and N=1 components. The nonlinear evolution during the VDE is strongly affected by the relative amplitude of the N=1 to the N=0 modes. (author)

Stationary Flowing Liquid Lithium (SFLiLi) systems for tokamaks

Science.gov (United States)

Zakharov, Leonid; Gentile, Charles; Roquemore, Lane

2013-10-01

The present approach to magnetic fusion which relies on high recycling plasma-wall interaction has exhausted itself at the level of TFTR, JET, JT-60 devices with no realistic path to the burning plasma. Instead, magnetic fusion needs a return to its original idea of insulation of the plasma from the wall, which was the dominant approach in the 1970s and upon implementations has a clear path to the DEMO device with PDT ~= 100 MW and Qelectric > 1 . The SFLiLi systems of this talk is the technology tool for implementation of the guiding idea of magnetic fusion. It utilizes the unique properties of flowing LiLi to pump plasma particles and, thus, insulate plasma from the walls. The necessary flow rate, ~= 1 g3/s, is very small, thus, making the use of lithium practical and consistent with safety requirements. The talk describes how chemical activity of LiLi, which is the major technology challenge of using LiLi in tokamaks, is addressed by SFLiLi systems at the level of already performed (HT-7) experiment, and in ongoing implementations for a prototype of SFLiLi for tokamak divertors and the mid-plane limiter for EAST tokamak (to be tested in the next experimental campaign). This work is supported by US DoE contract No. DE-AC02-09-CH11466.

Optimization of magnetic field system for glass spherical tokamak GLAST-III

International Nuclear Information System (INIS)

Ahmad, Zahoor; Ahmad, S; Naveed, M A; Deeba, F; Javeed, M Aqib; Batool, S; Hussain, S; Vorobyov, G M

2017-01-01

GLAST-III (Glass Spherical Tokamak) is a spherical tokamak with aspect ratio A = 2. The mapping of its magnetic system is performed to optimize the GLAST-III tokamak for plasma initiation using a Hall probe. Magnetic field from toroidal coils shows 1/ R dependence which is typical with spherical tokamaks. Toroidal field (TF) coils can produce 875 Gauss field, an essential requirement for electron cyclotron resonance assisted discharge. The central solenoid (CS) of GLAST-III is an air core solenoid and requires compensation coils to reduce unnecessary magnetic flux inside the vessel region. The vertical component of magnetic field from the CS in the vacuum vessel region is reduced to 1.15 Gauss kA −1 with the help of a differential loop. The CS of GLAST can produce flux change up to 68 mVs. Theoretical and experimental results are compared for the current waveform of TF coils using a combination of fast and slow capacitor banks. Also the magnetic field produced by poloidal field (PF) coils is compared with theoretically predicted values. It is found that calculated results are in good agreement with experimental measurement. Consequently magnetic field measurements are validated. A tokamak discharge with 2 kA plasma current and pulse length 1 ms is successfully produced using different sets of coils. (paper)

Magnetic field shielding system in a tokamak experimental power reactor (EPR): concept and calculations

International Nuclear Information System (INIS)

Peng, Y.K.M.; Marcus, F.B.; Dory, R.A.; Moore, J.R.

1975-01-01

A poloidal magnetic field shielding system is proposed for a tokamak EPR. This coil system minimizes the pulsed poloidal field that intersects the TF (toroidal field) coils and hence reduces the risk of superconductor quenching and structural failure of the coils. Based on an idealized shielding model, we have determined the configurations for the OH (ohmic heating), the S-VF (shield-vertical field), and the T-VF (trimming-vertical field) coils in a typical tokamak EPR. It is found that the pulsed poloidal field strength is greatly reduced in the TF coil region. The overall requirement in stored plasma and vertical field energy is also substantially reduced when compared with conventional EPR designs. Use of this field shielding system is expected to enhance reliability of the superconducting TF coils in a tokamak EPR

Slow bank system of SINP-Tokamak: A short report

International Nuclear Information System (INIS)

Ray, R.; Ranjan, P.; Chowdhury, S.; Bose, S.

1997-01-01

SINP Tokamak was made operational in July, 1987. The power supply system of the tokamak at that time was designed for a plasma duration of around 2 ms for a peak plasma current of 75 kA. Efforts were directed to increase this duration to 20 ms with the help of a slow bank system designed to work in conjunction with the original fast bank system. The design aspects of the system were completed and the system has been partially executed. Subsequent to this partial implementation, efforts were directed to incorporate the necessary control system and interface facilities between the existing fast bank and the developed slow bank systems. The significant features of the control circuits are that they work according to a well thought out sequences of logic and are designed to guard against possible failures in the existing or the developed power supplies. Efforts have been put to make the operation of the system as much user-friendly as could be worked out within certain practical constraints. The control circuit and interface facilities have been put to extensive tests and are found to work satisfactorily. The entire power supply system is now in active use for different research programmes in the group. (author)

Tokamak control simulator

International Nuclear Information System (INIS)

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

1976-01-01

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

Systematic design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak

NARCIS (Netherlands)

Hennen, B.A.; Westerhof, E.; Nuij, Pwjm; M.R. de Baar,; Steinbuch, M.

2012-01-01

Suppression of tearing modes is essential for the operation of tokamaks. This paper describes the design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak. The two main control tasks of this feedback control system are the radial alignment of electron

Pulse length assessment of compact ignition tokamak designs

International Nuclear Information System (INIS)

Stotler, D.P.; Pomphrey, N.

1989-07-01

A time-dependent zero-dimensional code has been developed to assess the pulse length and auxiliary heating requirements of Compact Ignition Tokamak (CIT) designs. By taking a global approach to the calculation, parametric studies can be easily performed. The accuracy of the procedure is tested by comparing with the Tokamak Simulation Code which uses theory-based thermal diffusivities. A series of runs is carried out at various levels of energy confinement for each of three possible CIT configurations. It is found that for cases of interest, ignition or an energy multiplication factor Q /approxreverse arrowgt/ 7 can be attained within the first half of the planned five-second flattop with 10--40 MW of auxiliary heating. These results are supported by analytic calculations. 18 refs., 7 figs., 2 tabs

DOUBLE code simulations of emissivities of fast neutrals for different plasma observation view-lines of neutral particle analyzers on the COMPASS tokamak

Science.gov (United States)

Mitosinkova, K.; Tomes, M.; Stockel, J.; Varju, J.; Stano, M.

2018-03-01

Neutral particle analyzers (NPA) measure line-integrated energy spectra of fast neutral atoms escaping the tokamak plasma, which are a product of charge-exchange (CX) collisions of plasma ions with background neutrals. They can observe variations in the ion temperature T i of non-thermal fast ions created by additional plasma heating. However, the plasma column which a fast atom has to pass through must be sufficiently short in comparison with the fast atom’s mean-free-path. Tokamak COMPASS is currently equipped with one NPA installed at a tangential mid-plane port. This orientation is optimal for observing non-thermal fast ions. However, in this configuration the signal at energies useful for T i derivation is lost in noise due to the too long fast atoms’ trajectories. Thus, a second NPA is planned to be connected for the purpose of measuring T i. We analyzed different possible view-lines (perpendicular mid-plane, tangential mid-plane, and top view) for the second NPA using the DOUBLE Monte-Carlo code and compared the results with the performance of the present NPA with tangential orientation. The DOUBLE code provides fast-atoms’ emissivity functions along the NPA view-line. The position of the median of these emissivity functions is related to the location from where the measured signal originates. Further, we compared the difference between the real central T i used as a DOUBLE code input and the T iCX derived from the exponential decay of simulated energy spectra. The advantages and disadvantages of each NPA location are discussed.

Hybrid model for simulation of plasma jet injection in tokamak

Science.gov (United States)

Galkin, Sergei A.; Bogatu, I. N.

2016-10-01

Hybrid kinetic model of plasma treats the ions as kinetic particles and the electrons as charge neutralizing massless fluid. The model is essentially applicable when most of the energy is concentrated in the ions rather than in the electrons, i.e. it is well suited for the high-density hyper-velocity C60 plasma jet. The hybrid model separates the slower ion time scale from the faster electron time scale, which becomes disregardable. That is why hybrid codes consistently outperform the traditional PIC codes in computational efficiency, still resolving kinetic ions effects. We discuss 2D hybrid model and code with exact energy conservation numerical algorithm and present some results of its application to simulation of C60 plasma jet penetration through tokamak-like magnetic barrier. We also examine the 3D model/code extension and its possible applications to tokamak and ionospheric plasmas. The work is supported in part by US DOE DE-SC0015776 Grant.

A systems analysis of the ARIES tokamak reactors

International Nuclear Information System (INIS)

Bathke, C.G.

1992-01-01

The multi-institutional ARIES study has completed a series of cost-of-electricity optimized conceptual designs of commercial tokamak fusion reactors that vary the assumed advances in technology and physics. A comparison of these designs indicates the cost benefit of various design options. A parametric systems analysis suggests a possible means to obtain a marginally competitive fusion reactor

Transients and burn dynamics in advanced tokamak fusion reactors

International Nuclear Information System (INIS)

Mantsinen, M.J.; Salomaa, R.R.E.

1994-01-01

Transient behavior of D 3 He-tokamak reactors is investigated numerically using a zero-dimensional code with prescribed profiles. Pure D 3 He start-up is compared to DT-assisted and DT-ignited start-ups. We have considered two categories of transients which could extinguish steady fusion burn: fuelling interruptions and sudden confinement changes similar to the L → H transients occurring in present-day tokamaks. Shutdown with various current and density ramp-down scenarios are studied, too. (author)

Research using small tokamaks

International Nuclear Information System (INIS)

1991-01-01

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

Modular pump limiter systems for large tokamaks

International Nuclear Information System (INIS)

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

1987-09-01

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

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)

Tokamak TCABR: Acquisition system, data analysis, and remote participation using MDSplus

International Nuclear Information System (INIS)

Sá, W.P. de

2012-01-01

Highlights: ► The implementation of MDSplus in TCABR tokamak. ► Interfaces between the system already installed and the MDSplus. - Abstract: Each plasma physics laboratory has a proprietary scheme to control and data acquisition system. Usually, it is different from one laboratory to another. It means that each laboratory has its own way to control the experiment and retrieving data from the database. Fusion research relies to a great extent on international collaboration and this private system makes it difficult to follow the work remotely. The TCABR data analysis and acquisition system has been upgraded to support a joint research programme using remote participation technologies. The choice of MDSplus (Model Driven System plus) is proved by the fact that it is widely utilized, and the scientists from different institutions may use the same system in different experiments in different tokamaks without the need to know how each system treats its acquisition system and data analysis. Another important point is the fact that the MDSplus has a library system that allows communication between different types of language (JAVA, Fortran, C, C++, Python) and programs such as MATLAB, IDL, OCTAVE. In the case of tokamak TCABR interfaces (object of this paper) between the system already in use and MDSplus were developed, instead of using the MDSplus at all stages, from the control, and data acquisition to the data analysis. This was done in the way to preserve a complex system already in operation and otherwise it would take a long time to migrate. This implementation also allows add new components using the MDSplus fully at all stages.

Tokamak TCABR: Acquisition system, data analysis, and remote participation using MDSplus

Energy Technology Data Exchange (ETDEWEB)

Sa, W.P. de, E-mail: pires@if.usp.br [Instituto de Fisica, Universidade de Sao Paulo, Rua do Matao, Travessa R, 187, CEP 05508-090 Cidade Universitaria, Sao Paulo (Brazil)

2012-12-15

Highlights: Black-Right-Pointing-Pointer The implementation of MDSplus in TCABR tokamak. Black-Right-Pointing-Pointer Interfaces between the system already installed and the MDSplus. - Abstract: Each plasma physics laboratory has a proprietary scheme to control and data acquisition system. Usually, it is different from one laboratory to another. It means that each laboratory has its own way to control the experiment and retrieving data from the database. Fusion research relies to a great extent on international collaboration and this private system makes it difficult to follow the work remotely. The TCABR data analysis and acquisition system has been upgraded to support a joint research programme using remote participation technologies. The choice of MDSplus (Model Driven System plus) is proved by the fact that it is widely utilized, and the scientists from different institutions may use the same system in different experiments in different tokamaks without the need to know how each system treats its acquisition system and data analysis. Another important point is the fact that the MDSplus has a library system that allows communication between different types of language (JAVA, Fortran, C, C++, Python) and programs such as MATLAB, IDL, OCTAVE. In the case of tokamak TCABR interfaces (object of this paper) between the system already in use and MDSplus were developed, instead of using the MDSplus at all stages, from the control, and data acquisition to the data analysis. This was done in the way to preserve a complex system already in operation and otherwise it would take a long time to migrate. This implementation also allows add new components using the MDSplus fully at all stages.

A modular system of codes for discharge data storage used at Tokamak TJ-1

International Nuclear Information System (INIS)

Guasp, J.

1983-01-01

The code system is able to sample the discharge data at a 83 KHsub(z) rate using a LPA-11K controller and AD-11K converters and store it on structured files. The file details are transparent to the retrieval and plotting programs. Graphs are available on both Tektronix 4010-1 display and Versatec 1200 plotter. The system allows data storing and retrieval on PDP-11/34 and UNIVAC-1100/8 computers as well as on magnetic tape. (author)

Including collisions in gyrokinetic tokamak and stellarator simulations

International Nuclear Information System (INIS)

Kauffmann, Karla

2012-01-01

Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the δf method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal ITGs and

Including collisions in gyrokinetic tokamak and stellarator simulations

Energy Technology Data Exchange (ETDEWEB)

Kauffmann, Karla

2012-04-10

Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the {delta}f method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal

3He functions in tokamak-pumped laser systems

International Nuclear Information System (INIS)

Jassby, D.L.

1986-10-01

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

Interactive exploration of tokamak turbulence simulations in virtual reality

International Nuclear Information System (INIS)

Kerbel, G.D.; Pierce, T.; Milovich, J.L.; Shumaker, D.E.

1996-01-01

We have developed an immersive visualization system designed for interactive data exploration as an integral part of our computing environment for studying tokamak turbulence. This system of codes can reproduce the results of simulations visually for scrutiny in real time, interactively and with more realism than ever before. At peak performance, the VR system can present for view some 400 coordinated images per second. The long term vision this approach targets is a open-quote holodeck-like close-quote virtual-reality environment in which one can explore gyrofluid or gyrokinetic plasma simulations interactively and in real time, visually, with concurrent simulations of experimental diagnostic devices. In principle, such a open-quote virtual tokamak close-quote computed environment could be as all encompassing or as focussed as one likes, in terms of the physics involved. The computing framework in one within which a group of researchers can work together to produce a real and identifiable product with easy access to all contributions. This could be our version of NASA's next generation Numerical Wind Tunnel. The principal purpose of this VR capability for Numerical Tokamak simulation is to provide interactive visual experience to help create new ways of understanding aspects of the convective transport processes operating in tokamak fusion experiments. The effectiveness of the visualization method is strongly dependent on the density of frame-to-frame correlation. Below a threshold of this quantity, short term visual memory does not bridge the gap between frames well enough for there to exist a strong visual connection. Above the threshold, evolving structures appear clearly. The visualizations show the 3D structure of vortex evolution and the gyrofluid motion associated with it. We discovered that it was very helpful for visualizing the cross field flows to compress the virtual world in the toroidal angle

Fokker-Planck code for the quasi-linear absorption of electron cyclotron waves in a tokamak plasma

International Nuclear Information System (INIS)

Meyer, R.L.; Giruzzi, G.; Krivenski, V.

1986-01-01

We present the solution of the kinetic equation describing the quasi-linear evolution of the electron momentum distribution function under the influence of the electron cyclotron wave absorption. Coulomb collisions and the dc electric field in a tokamak plasma. The solution of the quasi-linear equation is obtained numerically using a two-dimensional initial value code following an ADI scheme. Most emphasis is given to the full non-linear and self-consistent problem, namely, the wave amplitude is evaluated at any instant and any point in space according to the actual damping. This is necessary since wave damping is a very sensitive function of the slope of the local momentum distribution function because the resonance condition relates the electron momentum to the location of wave energy deposition. (orig.)

Long pulse neutral beam system for the Tokamak Physics Experiment

International Nuclear Information System (INIS)

Grisham, L.R.; Bowen, O.N.; Dahlgren, F.; Edwards, J.W.; Kamperschroer, J.; Newman, R.; O'Connor, T.; Ramakrishnan, S.; Rossi, G.; Stevenson, T.; Halle, A. von; Wright, K.E.

1995-01-01

The Tokamak Physics Experiment (TPX) is planned as a long-pulse or steady-state machine to serve as a successor to the Tokamak Fusion Test Reactor (TFTR). The neutral beam component of the heating and current drive systems will be provided by a TFTR beamline modified to allow operation for pulse lengths of 1000s. This paper presents a brief overview of the conceptual design which has been carried out to determine the changes to the beamline and power supply components that will be required to extend the pulse length from its present limitation of 1s at full power. The modified system, like the present one, will be capable of injecting about 8MW of power as neutral deuterium. The initial operation will be with a single beamline oriented co-directional to the plasma current, but the TPX system design is capable of accommodating an additional co-directional beamline and a counter-directional beamline. ((orig.))

Fast computational scheme for feedback control of high current fusion tokamaks

International Nuclear Information System (INIS)

Dong, J.Q.; Khayrutdinov, R.; Azizov, E.; Jardin, S.

1992-01-01

An accurate and fast numerical model of tokamak plasma evolution is presented. In this code (DINA) the equilibrium problem of plasmas with free boundaries in externally changing magnetic fields is solved simultaneously with the plasma transport equation. The circuit equations are solved for the vacuum vessel and passive and active coils. The code includes pellet injection, neutral beam heating, auxiliary heating, and alpha particle heating. Bootstrap and beam-driven plasma currents are accounted for. An inverse variable technique is utilized to obtain the coordinates of the equilibrium magnetic surfaces. This numerical algorithm permits to determine the flux coordinates very quickly and accurately. The authors show that using the fully resistive MHD analysis the region of stability (to vertical motions) is wider than using the rigid displacement model. Comparing plasma motions with the same gain, it is seen that the plasma oscillates more in the rigid analysis than in the MHD analysis. They study the influence of the pick up coil's location and the possibility of control of the plasma vertical position. They use a simple modification of the standard control law that enables the control of the plasma with pick up coils located at any position. This flexibility becomes critical in the design of future complex high current tokamak systems. The fully resistive MHD model permits to obtain accurate estimates of the plasma response. This approach yields computational time savings of one to two orders of magnitude with respect to other existing MHD models. In this sense, conventional numerical algorithms do not provide suitable models for application of modern control techniques into real time expert systems. The proposed inverse variable technique is rather suitable for incorporation in a comprehensive expert system for feedback control of fusion tokamaks in real time

The Thomson Scattering System on the Lithium Tokamak eXperiment (LTX)

International Nuclear Information System (INIS)

Strickler, T.; Majeski, R.; Kaita, R.; LeBlanc, B.

2008-01-01

The Lithium Tokamak eXperiment (LTX) is a spherical tokamak with R0 = 0.4m, a = 0.26m, BTF ∼ 3.4kG, IP ∼ 400kA, and pulse length ∼ 0.25s. The goal of LTX is to investigate tokamak plasmas that are almost entirely surrounded by a lithium-coated plasma-facing shell conformal to the last closed magnetic flux surface. Based on previous experimental results and simulation, it is expected that the low-recycling liquid lithium surfaces will result in higher temperatures at the plasma edge, flatter overall temperature profiles, centrally-peaked density profiles, and an increased confinement time. To test these predictions, the electron temperature and density profiles in LTX will be measured by a multi-point Thomson scattering system (TVTS). Initially, TS measurements will be made at up to 12 simultaneous points between the plasma center and plasma edge. Later, high resolution edge measurements will be deployed to study the lithium edge physics in greater detail. Technical challenges to implementing the TS system included limited 'line of sight' access to the plasma due to the plasma-facing shell and problems associated with the presence of liquid lithium.

Summary report on tokamak confinement experiments

International Nuclear Information System (INIS)

1982-03-01

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

Overview of tritium systems for the Compact Ignition Tokamak

International Nuclear Information System (INIS)

Bartlit, J.R.; Gruetzmacher, K.M.; Fleming, R.B.

1987-01-01

The Compact Ignition Tokamak (CIT) is being designed at several laboratories to produce and study fully ignited plasma discharges. The tritium systems which will be needed for CIT include fueling systems and radiation monitoring and safety systems. Design of the tritium systems is the responsibility of the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. Major new tritium systems for CIT include a pellet injector, an air detritiation system and a glovebox atmosphere detritiation system. The pellet injector is being developed at Oak Ridge National Laboratory. 7 refs., 2 figs

Tokamak

International Nuclear Information System (INIS)

Wesson, John.

1996-01-01

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

The impurity transport in HT-6M tokamak

International Nuclear Information System (INIS)

Xu Wei; Wan Baonian; Xie Jikang

2003-01-01

The space-time profile of impurities has been measured with a multichannel visible spectroscopic detect system and UV rotation-mirror system in the HT-6M tokamak. An ideal impurity transport code has been used to simulate impurities (carbon and oxygen) behaviour during the OHM discharge. The profiles of impurities diffusion and convection coefficient, impurities ion densities in different ionized state, loss power density and effective charge number have been derived. The impurity behaviour during low-hybrid current drive has also been analyzed, the results show that the confinement of particles, impurities and energy has been improved, and emission power and effective charge number have been reduced

Track-mounted remote handling system for the Tokamak Fusion Engineering Test

International Nuclear Information System (INIS)

Kelly, V.P.; Berger, J.D.; Daubert, R.L.; Yount, J.A.

1982-01-01

Concepts for remote handling machines (IVM) designed to transverse the interior of toroidal vessels with guidance and support from track systems have been developed for the proposed Tokamak Fusion Engineering Test (TFET). TFET has been proposed as an upgrade for the Tokamak Fusion Test Reactor (TFTR), currently nearing completion. The track-mounted IVMs were conceived to perform in-vessel remote maintenance for TFET, including removal and replacement of pump limiter blades and protective tiles as well as other maintenance-related tasks such as vessel wall inspection leak testing and interior cleanup. The conceptual IVMs consist of three manipulator arms mounted on a common frame member: a single power manipulator arm with high load carrying capacity and two lower-capacity servomanipulator arms. Descriptions of the IVM concepts, in-vessel track systems, and ex-vessel handling systems are presented

Sawtooth driven particle transport in tokamak plasmas

International Nuclear Information System (INIS)

Nicolas, T.

2013-01-01

The radial transport of particles in tokamaks is one of the most stringent issues faced by the magnetic confinement fusion community, because the fusion power is proportional to the square of the pressure, and also because accumulation of heavy impurities in the core leads to important power losses which can lead to a 'radiative collapse'. Sawteeth and the associated periodic redistribution of the core quantities can significantly impact the radial transport of electrons and impurities. In this thesis, we perform numerical simulations of sawteeth using a nonlinear tridimensional magnetohydrodynamic code called XTOR-2F to study the particle transport induced by sawtooth crashes. We show that the code recovers, after the crash, the fine structures of electron density that are observed with fast-sweeping reflectometry on the JET and TS tokamaks. The presence of these structure may indicate a low efficiency of the sawtooth in expelling the impurities from the core. However, applying the same code to impurity profiles, we show that the redistribution is quantitatively similar to that predicted by Kadomtsev's model, which could not be predicted a priori. Hence finally the sawtooth flushing is efficient in expelling impurities from the core. (author) [fr

TOPIC: a debugging code for torus geometry input data of Monte Carlo transport code

International Nuclear Information System (INIS)

Iida, Hiromasa; Kawasaki, Hiromitsu.

1979-06-01

TOPIC has been developed for debugging geometry input data of the Monte Carlo transport code. the code has the following features: (1) It debugs the geometry input data of not only MORSE-GG but also MORSE-I capable of treating torus geometry. (2) Its calculation results are shown in figures drawn by Plotter or COM, and the regions not defined or doubly defined are easily detected. (3) It finds a multitude of input data errors in a single run. (4) The input data required in this code are few, so that it is readily usable in a time sharing system of FACOM 230-60/75 computer. Example TOPIC calculations in design study of tokamak fusion reactors (JXFR, INTOR-J) are presented. (author)

Comprehensive numerical modelling of tokamaks

International Nuclear Information System (INIS)

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

1991-01-01

We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell

Preliminary project of s Thomson scattering system for the ETE tokamak

International Nuclear Information System (INIS)

Berni, Luiz Angelo

1997-01-01

This report presents the preliminary project of the injection and laser light block system for the Thomson (ET) scattering diagnostic to be implanted at the ETE spheric tokamak of the Instituto Nacional de Pesquisas Espaciais (INPE/LAP). Also, a scanning system for the optics of scattered light

Optimization study of normal conductor tokamak for commercial neutron source

Science.gov (United States)

Fujita, T.; Sakai, R.; Okamoto, A.

2017-05-01

The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

Overvoltage protection for magnetic system during disruption in tokamak

International Nuclear Information System (INIS)

Zhang, Ming; Li, Xiaolong; He, Yang; Zhang, Jun; Chen, Zhongyong; Yu, Kexun

2015-01-01

Highlights: • We investigate the way to limit the plasma disruption overvoltage by using the MOVs. • An overvoltage model of plasma disruption is introduced. • The overvoltage protection scheme has been verified by disruption experiments. • The overvoltage during plasma disruption can be limited to 330 V. - Abstract: During a plasma disruption the magnetic flux in the tokamak changes rapidly, which in most cases will cause high-voltage surges among the magnetic systems and may bring severe damage to the components if there is no overvoltage protection. This paper investigates the way to limit the plasma disruption overvoltage and absorb the energy with the use of metal oxide varistors (MOVs). An overvoltage model of plasma disruption is introduced which can be used for the simulation of plasma disruption and the analysis of the overvoltage. The effectiveness of the overvoltage protection system is validated with disruption experiments. It shows that by optimizing the varistors voltage, the overvoltage during plasma disruption can be limited to an ideal low value. Now the overvoltage protection system has been deployed in J-TEXT tokamak and serves well for daily experiments.

Energy balance in TM-1-MH Tokamak (ohmical heating)

Science.gov (United States)

Stoeckel, J.; Koerbel, S.; Kryska, L.; Kopecky, V.; Dadalec, V.; Datlov, J.; Jakubka, K.; Magula, P.; Zacek, F.; Pereverzev, G. V.

1981-10-01

Plasma in the TM-1-MH Tokamak was experimentally studied in the parameter range: tor. mg. field B = 1,3 T, plasma current I sub p = 14 kA, electron density N sub E 3.10 to the 19th power cubic meters. The two numerical codes are available for the comparison with experimental data. TOKATA-code solves simplified energy balance equations for electron and ion components. TOKSAS-code solves the detailed energy balance of the ion component.

Spectroscopic system for impurity measurements in the TJ-1 Tokamak of JEN

International Nuclear Information System (INIS)

Navas, G.; Zurro, B.

1982-01-01

we describe a spectroscopic system with spatial resolution capability that has been configured for plasma diagnostic in the TJ-1 Tokamak of JEN. The experimental system, based on a one meter monochromator, has been absolutely calibrated using a tungsten-halogen lamp. The calibration procedures and the absolute spectral sensitivity are presented as well as its dependence with the polarization. A simplified spectroscopic model of the radiation emitted by the intrinsic plasma impurities (C, 0, . . . ) has been developed. A one dimensional model of the temporal evolution of various ionization stages in coronal equilibrium is used to predict the electron temperature and impurity concentration. This model has been applied to experimental data from several Tokamaks. (Author) 23 refs

Fast reciprocating probe system on the EAST superconducting tokamak

Energy Technology Data Exchange (ETDEWEB)

Zhang, W.; Chang, J. F.; Wan, B. N.; Xu, G. S.; Li, B.; Xu, C. S.; Yan, N.; Wang, L.; Liu, S. C.; Jiang, M.; Liu, P. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031 (China); Xiao, C. J. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031 (China); Department of Physics and Engineering Physics, Plasma Physics Laboratory, University of Saskatchewan, Saskatoon SK S7N 5E2 (Canada)

2010-11-15

A new fast reciprocating probe system (FRPS) has been built and installed on the outer midplane of the EAST tokamak to investigate the profiles of the boundary plasma parameters such as electron density and temperature. The system consists of a two-stage motion drive mechanism: slow motion and fast motion. The fast motion is powered by a servo motor, which drives the probe horizontally up to 50 cm to scan the edge region of the EAST tokamak. The maximum velocity achieved is 2 m/s. High velocity and flexible control of the fast motion are the remarkable features of this FRPS. A specially designed connector installed at the front end of the probe shaft makes it easy to install or replace the probe head on FRPS. During the latest experimental campaign in the spring of 2010, a probe head with seven tips, including two tips for a Mach probe, has been used. An example is given for simultaneous profile measurements of the plasma temperature, plasma density, and the plasma flow velocity.

Review of the equilibrium fitting for non-circular tokamak

International Nuclear Information System (INIS)

Luo Jiarong

2002-01-01

As the equilibrium fitting code (EFIT) is developing to perform the magnetic and the kinetic-magnetic analysis for tokamak device operation, it can be not only run in either the fitting mode or the equilibrium mode but also control operation of modern experimental fusion device. The history of EFIT code and its capabilities are described in section 2. A brief description of the off-line EFIT code and the development of the real-time EFIT (RTEFIT) code is shown in section 3 and 4 respectively. In the last section the summary is given

The ARIES-I tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

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

Application of advanced composites in tokamak magnet systems

International Nuclear Information System (INIS)

Long, C.J.

1977-11-01

The use of advanced (high-modulus) composites in superconducting magnets for tokamak fusion reactors is discussed. The most prominent potential application is as the structure in the pulsed poloidal-field coil system, where a significant reduction in eddy currents could be achieved. Present low-temperature data on the advanced composites are reviewed briefly; they are too meager to do more than suggest a broad class of composites for a particular application

Benchmark studies of BOUT++ code and TPSMBI code on neutral transport during SMBI

Energy Technology Data Exchange (ETDEWEB)

Wang, Y.H. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Z.H., E-mail: zhwang@swip.ac.cn [Southwestern Institute of Physics, Chengdu 610041 (China); Guo, W., E-mail: wfguo@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China); Ren, Q.L. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sun, A.P.; Xu, M.; Wang, A.K. [Southwestern Institute of Physics, Chengdu 610041 (China); Xiang, N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China)

2017-06-09

SMBI (supersonic molecule beam injection) plays an important role in tokamak plasma fuelling, density control and ELM mitigation in magnetic confinement plasma physics, which has been widely used in many tokamaks. The trans-neut module of BOUT++ code is the only large-scale parallel 3D fluid code used to simulate the SMBI fueling process, while the TPSMBI (transport of supersonic molecule beam injection) code is a recent developed 1D fluid code of SMBI. In order to find a method to increase SMBI fueling efficiency in H-mode plasma, especially for ITER, it is significant to first verify the codes. The benchmark study between the trans-neut module of BOUT++ code and the TPSMBI code on radial transport dynamics of neutral during SMBI has been first successfully achieved in both slab and cylindrical coordinates. The simulation results from the trans-neut module of BOUT++ code and TPSMBI code are consistent very well with each other. Different upwind schemes have been compared to deal with the sharp gradient front region during the inward propagation of SMBI for the code stability. The influence of the WENO3 (weighted essentially non-oscillatory) and the third order upwind schemes on the benchmark results has also been discussed. - Highlights: • A 1D model of SMBI has developed. • Benchmarks of BOUT++ and TPSMBI codes have first been finished. • The influence of the WENO3 and the third order upwind schemes on the benchmark results has also been discussed.

Tokamak Fusion Core Experiment (TFCX) special-purpose remote maintenance systems

International Nuclear Information System (INIS)

Masson, L.S.; Welland, H.J.

1985-01-01

A key element in the preconceptual design of the Tokamak Fusion Core Experiment (TFCX) was the development of design concepts for special-purpose remote maintenance systems. Included were systems for shield sector replacement, vacuum vessel sector and toroidal field coil replacement, limiter blade replacement, protective tile replacement, and general-purpose maintenance. This paper addresses these systems as they apply to the copper toroidal field (TF) coil version of the TFCX

Interlock System for the COMPASS Tokamak

Energy Technology Data Exchange (ETDEWEB)

Hron, M.; Adamek, J.; Pisacka, J.; Panek, R. [Institute of Plasma Physics AS CR, Association EURATOM/IPP.CR, Prague (Czech Republic); Sova, J.; Siba, J.; Kovarj, J. [Department of Control Engineering, Faculty of Electrical Engineering, Prague (Czech Republic)

2009-07-01

Full text of publication follows: The COMPASS tokamak (R=0.56 m, a=0.18 - 0.23 m) is starting operation presently at Institute of Plasma Physics AS CR in Prague. An important issue of the operation is the safety of the personnel and machine protection against faults, presented in this contribution. The personnel protection is based on a restricted access into the experimental hall during the operation of potentially dangerous systems. A tokamak hall access system, based on Honeywell WIN-PAK (tm) 2005, allows to set the status of the experimental area (open/closed) and to control the in and out movement of persons using access cards. On top of this, a check of the whole experimental area by the operator is enforced before the hall enclosure. A hardware interlock then interprets the experimental hall status and controls the operation of key systems accordingly. The permit for operation is granted and the real status of the systems is reported by hard wired potential less contacts. The control procedure is based on a PLC MicroPEL M66. This PLC provides HW interface between Actuators (Relays and switches) and it is connected on PESNET bus. Its programming is done using language Simple v.2 in Winstudio IDE. Second site of personnel protection system is created on PC where runs a .NET application on MSWindows XP or 2000. This PC is connected with PLC via PESNET bus (on RS485 layer) and it generates all control signals to PLC from the operator. Simultaneously, the PC receives all warning and alarm signals from the PLC. This signals are displayed on a screen of the PC in real-time, this way the GUI provides visualization of the controlled process. Except for this fact the operator is informed about the status of the system and individual subsystems on a PC via an operator's panel. Further we will describe the machine protection which uses similar system for checking conditions for the start of a shot. Fast key processes which have to be checked during the shot are

Two dimensional code for modeling of high ione cyclotron harmonic fast wave heating and current drive

International Nuclear Information System (INIS)

Grekov, D.; Kasilov, S.; Kernbichler, W.

2016-01-01

A two dimensional numerical code for computation of the electromagnetic field of a fast magnetosonic wave in a tokamak at high harmonics of the ion cyclotron frequency has been developed. The code computes the finite difference solution of Maxwell equations for separate toroidal harmonics making use of the toroidal symmetry of tokamak plasmas. The proper boundary conditions are prescribed at the realistic tokamak vessel. The currents in the RF antenna are specified externally and then used in Ampere law. The main poloidal tokamak magnetic field and the ''kinetic'' part of the dielectric permeability tensor are treated iteratively. The code has been verified against known analytical solutions and first calculations of current drive in the spherical torus are presented.

Multiple view fan beam polarimetry on Tokamak devices

International Nuclear Information System (INIS)

Geck, W.R.; Domier, C.W.; Luhmann, N.C.

1997-01-01

A polarimeter diagnostic is under development which utilizes several fan beams to accumulate line integrated Faraday rotation data in a Tokamak plasma. The utilization of a fan beam configuration over that of conventional vertical view polarimeter systems significantly reduces access requirements. The high angular separation inherent in a fan beam implementation increases plasma coverage and eliminates the necessity of assumed plasma symmetries to generate high quality current density profiles. Codes have been developed to generate these high-resolution two-dimensional images of the plasma current profile from data collected at arbitrary positions and viewing angles. copyright 1997 American Institute of Physics

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

Energy Technology Data Exchange (ETDEWEB)

Jassby, D.L.

1986-10-01

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

A lithium deposition system for tokamak devices*

Science.gov (United States)

Graziul, Christopher; Majeski, Richard; Kaita, Robert; Hoffman, Daniel; Timberlake, John; Card, David

2002-11-01

The production of a lithium deposition system using commercially available components is discussed. This system is intended to provide a fresh lithium wall coating between discharges in a tokamak. For this purpose, a film 100-200 Å thick is sufficient to ensure that the plasma interacts solely with the lithium. A test system consisting of a lithium evaporator and a deposition monitor has been designed and constructed to investigate deposition rates and coverage. A Thermionics 3kW e-gun is used to rapidly evaporate small amounts of solid lithium. An Inficon XTM/2 quartz deposition monitor then measures deposition rate at varying distances, positions and angles relative to the e-gun crucible. Initial results from the test system will be presented. *Supported by US DOE contract #DE-AC02-76CH-03073

Studies of feedback stabilization of axisymmetric modes in deformable tokamak plasmas

International Nuclear Information System (INIS)

Ward, D.J.

1991-01-01

A new linear MHD stability code, NOVA-W, is described and applied to the study of the feedback stabilization of the axisymmetric mode in deformable tokamak plasma. The NOVA-W code is a modification of the non-variational MHD stability code NOVA that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The code has been tested for the case of passive stabilization against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. The NOVA-W code is used to examine the effects of plasma deformability on feedback stabilization. It is seen that plasmas with shaped cross sections have unstable motion different from a rigid shift. Plasma equilibria with large triangularity show particularly significant deviations from a uniform rigid shift. Furthermore, the placement of passive conductors is shown to modify the non-rigid components of the motion in a way that reduces the stabilizing effects of these conductors. The eigenfunction is also modified under the effects of active feedback. This deformation is seen to depend strongly on the position of the flux loops. These non-rigid components of the eigenfunction always serve to reduce the stabilizing effect of the active feedback system by reducing the measurable poloidal flux at the flux-loop locations

39GHz ECRH system for breakdown studies on the TCA tokamak

International Nuclear Information System (INIS)

Pochelon, A.; Goodman, T.; Whaley, D.; Tran, M.Q.; Reinhard, D.; Perrenoud, A.; Joedicke, B.; Mathews, H.G.; Kasparek, W.; Thumm, M.

1990-01-01

The design construction and first operation of a 39GHz ECRH system (300 kW, 100 ms) for low loop-voltage breakdown and startup-assist experiments on the TCA tokamak is described. (author) 5 refs., 6 figs., 1 tab

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

Energy Technology Data Exchange (ETDEWEB)

Sa, Wanderley Pires de

1996-12-31

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

Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

1993-01-01

The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility

2D full-wave simulation of waves in space and tokamak plasmas

Directory of Open Access Journals (Sweden)

Kim Eun-Hwa

2017-01-01

Full Text Available Simulation results using a 2D full-wave code (FW2D for space and NSTX fusion plasmas are presented. The FW2D code solves the cold plasma wave equations using the finite element method. The wave code has been successfully applied to describe low frequency waves in planetary magnetospheres (i.e., dipole geometry and the results include generation and propagation of externally driven ultra-low frequency waves via mode conversion at Mercury and mode coupling, refraction and reflection of internally driven field-aligned propagating left-handed electromagnetic ion cyclotron (EMIC waves at Earth. In this paper, global structure of linearly polarized EMIC waves is examined and the result shows such resonant wave modes can be localized near the equatorial plane. We also adopt the FW2D code to tokamak geometry and examine radio frequency (RF waves in the scape-off layer (SOL of tokamaks. By adopting the rectangular and limiter boundary, we compare the results with existing AORSA simulations. The FW2D code results for the high harmonic fast wave heating case on NSTX with a rectangular vessel boundary shows excellent agreement with the AORSA code.

2D full-wave simulation of waves in space and tokamak plasmas

Science.gov (United States)

Kim, Eun-Hwa; Bertelli, Nicola; Johnson, Jay; Valeo, Ernest; Hosea, Joel

2017-10-01

Simulation results using a 2D full-wave code (FW2D) for space and NSTX fusion plasmas are presented. The FW2D code solves the cold plasma wave equations using the finite element method. The wave code has been successfully applied to describe low frequency waves in planetary magnetospheres (i.e., dipole geometry) and the results include generation and propagation of externally driven ultra-low frequency waves via mode conversion at Mercury and mode coupling, refraction and reflection of internally driven field-aligned propagating left-handed electromagnetic ion cyclotron (EMIC) waves at Earth. In this paper, global structure of linearly polarized EMIC waves is examined and the result shows such resonant wave modes can be localized near the equatorial plane. We also adopt the FW2D code to tokamak geometry and examine radio frequency (RF) waves in the scape-off layer (SOL) of tokamaks. By adopting the rectangular and limiter boundary, we compare the results with existing AORSA simulations. The FW2D code results for the high harmonic fast wave heating case on NSTX with a rectangular vessel boundary shows excellent agreement with the AORSA code.

The critical temperature gradient model of plasma transport: applications to Jet and future tokamaks

International Nuclear Information System (INIS)

Rebut, P.H.; Lallia, P.P.; Watkins, M.L.

1989-01-01

The diversity and complexity of behaviour in tokamak plasmas place strong constraints on any model attempting a description in terms of a single underlying phenomenon. Assuming that turbulence in the magnetic topology is the underlying phenomenon, specific expressions for electron and ion heat flux are derived from heuristic and dimensional arguments. When used in plasma transport codes, rather satisfactory simulations of experimental results are achieved in different sized tokamaks in various regimes of operation. Predictions are given for the expected performance of JET at full planned power and implications for next step tokamaks are indicated

Thomson scattering on the PRETEXT Tokamak

International Nuclear Information System (INIS)

McCool, S.C.

1982-03-01

Ruby laser Thomson scattering was performed on the PRETEXT tokamak. A 10 Joule Q-switched laser and a 1 meter 10 channel polychromator were used to diagnose the electron temperature and density profiles in the PRETEXT plasma. These parameters were measured as a function of time and radial position on a shot to shot basis. The density measurement was calibrated by Rayleigh and Raman scattering and by comparison with data from a 4 mm microwave interferometer. Electron densities ranging from 1 x 10 12 cm -3 to 2 x 10 13 cm -3 and temperatures ranging from 3 eV to 400 eV were observed. Detailed measurements were made throughout the 40 ms discharge with particular emphasis on the current rise phase. The Thomson scattering data was used as input to a one dimensional magnetic diffusion code. This code modelled the evolution of the current density and safety factor profiles. The results of this analysis were compared with existing theories of tokamak current penetration. The growth of resitive MHD tearing modes was proposed as a likely explanation for the anomalously rapid current penetration observed in PRETEXT

Active cooling system for Tokamak in-vessel operation manipulator

Energy Technology Data Exchange (ETDEWEB)

Yuan, Jianjun, E-mail: yuanjj@sjtu.edu.cn; Chen, Tan; Li, Fashe; Zhang, Weijun; Du, Liang

2015-10-15

Highlights: • We summarized most of the challenges of fusion devices to robot systems. • Propose an active cooling system to protect all of the necessary components. • Trial design test and theoretical analysis were conducted. • Overall implementation of the active cooling system was demonstrated. - Abstract: In-vessel operation/inspection is an indispensable task for Tokamak experimental reactor, for a robot/manipulator is more capable in doing this than human being with more precise motion and less risk of damaging the ambient equipment. Considering the demanding conditions of Tokamak, the manipulator should be adaptable to rapid response in the extreme conditions such as high temperature, vacuum and so on. In this paper, we propose an active cooling system embedded into such manipulator. Cameras, motors, gearboxes, sensors, and other mechanical/electrical components could then be designed under ordinary conditions. The cooling system cannot only be a thermal shield since the components are also heat sources in dynamics. We carry out a trial test to verify our proposal, and analyze the active cooling system theoretically, which gives a direction on the optimization by varying design parameters, components and distribution. And based on thermal sensors monitoring and water flow adjusting a closed-loop feedback control of temperature is added to the system. With the preliminary results, we believe that the proposal gives a way to robust and inexpensive design in extreme environment. Further work will concentrate on overall implementation and evaluation of this cooling system with the whole inspection manipulator.

Economic evaluation of tokamak power plants

International Nuclear Information System (INIS)

Reid, R.L.; Steiner, D.

1977-01-01

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

Li-BES detection system for plasma turbulence measurements on the COMPASS tokamak

Energy Technology Data Exchange (ETDEWEB)

Berta, M. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Széchenyi István University, Győr (Hungary); Anda, G.; Bencze, A.; Dunai, D. [Wigner – RCP, HAS, Budapest (Hungary); Háček, P., E-mail: hacek@ipp.cas.cz [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Hron, M. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Kovácsik, A. [Wigner – RCP, HAS, Budapest (Hungary); Department of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Krbec, J. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Pánek, R. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Réfy, D.; Veres, G. [Wigner – RCP, HAS, Budapest (Hungary); Weinzettl, V. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Zoletnik, S. [Wigner – RCP, HAS, Budapest (Hungary)

2015-10-15

Highlights: • Li-BES detection system on the COMPASS tokamak is optimized observation system with high temporal resolution. • High sensitivity to low level light fluctuations. • Optics and detectors with electronics are placed in thermally stabilized compact box. • Fast deflection system allows us to measure background corrected electron density profiles on microsecond time-scale. - Abstract: A new Li beam emission spectroscopy (Li-BES) diagnostic system with a ∼ cm spatial resolution, and with beam energy ranging from 10 keV up to 120 keV and a 18 channel Avalanche photo diode (APD) detector system sampled at 2 MHz has been recently installed and tested on the COMPASS tokamak. This diagnostic allows to reconstruct density profile based on directly measured light profiles, and to follow turbulent behaviour of the edge plasma. The paper reports technical capabilities of this new system designed for fine spatio-temporal measurements of plasma electron density. Focusing on turbulence-induced fluctuation measurements, we demonstrate how physically relevant information can be extracted using the COMPASS Li-BES system.

Computation of tokamak equilibria with steady flow

International Nuclear Information System (INIS)

Kerner, W.; Tokuda, Shinji

1987-08-01

The equations for ideal MHD equilibria with stationary flow are reexamined and addressed as numerically applied to tokamak configurations with a free plasma boundary. Both the isothermal (purely toroidal flow) and the poloidal flow cases are treated. Experiment-relevant states with steady flow (so far only in the toroidal direction) are computed by the modified SELENE40 code. (author)

Continuous tokamaks

International Nuclear Information System (INIS)

Peng, Y.K.M.

1978-04-01

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

Development of atomic beam probe for tokamaks

Energy Technology Data Exchange (ETDEWEB)

Berta, M., E-mail: bertam@sze.hu [Széchenyi István University, EURATOM Association, Győr (Hungary); Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, EURATOM Association, Budapest (Hungary); Havlícek, J.; Háček, P. [Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics (Czech Republic)

2013-11-15

Highlights: • ABP is newly developed diagnostic. • Unique measurement method for the determination of plasma edge current variations caused by different transient events such as ELMs. • The design process has been fruitfully supported by the physically motivated computer simulations. • Li-BES system has been modified accordingly to the needs of the ABP. -- Abstract: The concept and development of a new detection method for light alkali ions stemming from diagnostic beams installed on medium size tokamak is described. The method allows us the simultaneous measurement of plasma density fluctuations and fast variations in poloidal magnetic field, therefore one can infer the fast changes in edge plasma current. The concept has been worked out and the whole design process has been done at Wigner RCP. The test detector with appropriate mechanics and electronics is already installed on COMPASS tokamak. General ion trajectory calculation code (ABPIons) has also been developed. Detailed calculations show the possibility of reconstruction of edge plasma current density profile changes with high temporal resolution, and the possibility of density profile reconstruction with better spatial resolution compared to standard Li-BES measurement, this is important for pedestal studies.

Development of atomic beam probe for tokamaks

International Nuclear Information System (INIS)

Berta, M.; Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S.; Havlícek, J.; Háček, P.

2013-01-01

Highlights: • ABP is newly developed diagnostic. • Unique measurement method for the determination of plasma edge current variations caused by different transient events such as ELMs. • The design process has been fruitfully supported by the physically motivated computer simulations. • Li-BES system has been modified accordingly to the needs of the ABP. -- Abstract: The concept and development of a new detection method for light alkali ions stemming from diagnostic beams installed on medium size tokamak is described. The method allows us the simultaneous measurement of plasma density fluctuations and fast variations in poloidal magnetic field, therefore one can infer the fast changes in edge plasma current. The concept has been worked out and the whole design process has been done at Wigner RCP. The test detector with appropriate mechanics and electronics is already installed on COMPASS tokamak. General ion trajectory calculation code (ABPIons) has also been developed. Detailed calculations show the possibility of reconstruction of edge plasma current density profile changes with high temporal resolution, and the possibility of density profile reconstruction with better spatial resolution compared to standard Li-BES measurement, this is important for pedestal studies

Effects of impurities and magnetic divertors on high-temperature tokamaks

International Nuclear Information System (INIS)

Meade, D.M.; Furth, H.P.; Rutherford, P.H.; Seidl, F.G.P.; Duechs, D.F.

1974-10-01

A one-dimensional tokamak plasma transport code has been adapted to include impurity influx, stripping, radiation, and diffusion, as well as the usual processes of hydrogen plasma and heat transport, recycling at the boundary, and multigeneration charge-exchange. Neutral-beam heating, adiabatic compression, and divertor boundary conditions are included as optional features. Illustrative computations are given for present-day and next-generation tokamaks. The problems of impurity control are discussed, and two technical approaches are examined in greater detail: the transient cold-plasma shield, and the poloidal divertor. (auth)

Advanced commercial tokamak study

International Nuclear Information System (INIS)

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

1985-12-01

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

Tokamak power systems studies at ANL

International Nuclear Information System (INIS)

Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.

1986-01-01

A number of advances in plasma physics and engineering promise to greatly improve the reactor prospects of tokamaks. The following features, in particular, are examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 7 T; (c) low toroidal current (I ≅ 5MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields produced in the plasma. In addition to matching desirable high-beta equilibria, this method is capable of producing a large variety of new equilibria, many of which look attractive. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors

Data bank system in JFT-2M tokamak

International Nuclear Information System (INIS)

Takada, S.; Matsuda, T.; Miura, Y.; Mori, M.; Kawakami, T.; Matoba, T.

1987-01-01

It is very important to keep suitably and use effectively experimental data in the field of fusion research. The data bank system for fusion experiment can be classified into the following two forms, according to the type of their use: (1) Rapid Analysis Data Bank System; (2) High Quality Data Bank System. And their features are summarized. The data bank system of the former type was prepared for JFT-2M tokamak on the basis of the above classification. By introduction of this data bank system, the following results can be obtained: (1) Rationalization of data analyzing procedure; (2) Improvement of reliability by exclusion of bad data; (3) Easy expansions of analyzing function and tool development; (4) Space saving by extraction and compression of key information

Tokamak Physics Experiment (TPX) power supply design and development

International Nuclear Information System (INIS)

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

1995-01-01

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

First experimental results with the Current Limit Avoidance System at the JET tokamak

Energy Technology Data Exchange (ETDEWEB)

De Tommasi, G. [Associazione EURATOM-ENEA-CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Galeani, S. [Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Rome (Italy); Jachmich, S. [Association EURATOM-Belgian State, Koninklijke Militaire School - Ecole Royale Militaire, B-1000 Brussels (Belgium); Joffrin, E. [IRFM-CEA, Centre de Cadarache, 13108 Saint-paul-lez-Durance (France); Lennholm, M. [EFDA Close Support Unit, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); European Commission, B-1049 Brussels (Belgium); Lomas, P.J. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Neto, A.C. [Associazione EURATOM-IST, Instituto de Plasmas e Fusao Nuclear, IST, 1049-001 Lisboa (Portugal); Maviglia, F. [Associazione EURATOM-ENEA-CREATE, Via Claudio 21, 80125 Napoli (Italy); McCullen, P. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Pironti, A. [Associazione EURATOM-ENEA-CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Rimini, F.G. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Sips, A.C.C. [European Commission, B-1049 Brussels (Belgium); Varano, G.; Vitelli, R. [Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Rome (Italy); Zaccarian, L. [CNRS, LAAS, 7 Avenue du Colonel Roche, F-31400 Toulouse (France); Universitè de Toulouse, LAAS, F-31400 Toulouse (France)

2013-06-15

The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.

Simulation of lower hybrid current drive in enhanced reversed shear plasmas in the tokamak fusion test reactor using the lower hybrid simulation code

International Nuclear Information System (INIS)

Kaita, R.; Bernabei, S.; Budny, R.

1996-01-01

The Enhanced Reversed Shear (ERS) mode has already shown great potential for improving the performance of the Tokamak Fusion Test Reactor (TFTR) and other devices. Sustaining the ERS, however, remains an outstanding problem. Lower hybrid (LH) current drive is a possible method for modifying the current profile and controlling its time evolution. To predict its effectiveness in TFTR, the Lower Hybrid Simulation Code (LSC) model is used in the TRANSP code and the Tokamak Simulation Code (TSC). Among the results from the simulations are the following. (1) Single-pass absorption is expected in TFTR ERS plasmas. The simulations show that the LH current follows isotherms of the electron temperature. The ability to control the location of the minimum in the q profile (q min ) has been demonstrated by varying the phase velocity of the launched LH waves and observing the change in the damping location. (2) LH current drive can been used to sustain the q min location. The tendency of qmin to drift inward, as the inductive current diffuses during the formation phase of the reversed shear discharge, is prevented by the LH current driven at a fixed radial location. If this results in an expanded plasma volume with improved confinement as high power neutral beam injection is applied, the high bootstrap currents induced during this phase can then maintain the larger qmin radius. (3) There should be no LH wave damping on energetic beam particles. The values of perpendicular index of refraction in the calculations never exceed about 20, while ions at TFR injection energies are resonant with waves having values closer to 100. Other issues being addressed in the study include the LH current drive efficiency in the presence of high bootstrap currents, and the effect of fast electron diffusion on LH current localization

Conceptual design of HL-2M tokamak control system

International Nuclear Information System (INIS)

Xia Fan; Chen Liaoyuan; Song Xianming; Zhang Jinhua; Lou Cuiwen; Pan Yudong

2009-01-01

The static architecture, dynamic behavior, control theory and simulation of HL-2M tokamak control system are described. The real-time network will be build for the communication of real-time control among its subsystems and universal timing system will be build to guarantee the synchronization among the subsystems. The duty to achieve preprogrammed parameters is carried out by plasma discharge control. In order to reduce the damage made by discharge exception, the error-handing mechanism of supervision system is considered. The controllers of magnetic control system are designed to control the current, shape and position of plasma and simulation system is designed for testing the controllers. (authors)

COMPREHENSIVE GYROKINETIC SIMULATION OF TOKAMAK TURBULENCE AT FINITE RELATIVE GYRORADIUS

International Nuclear Information System (INIS)

WALTZ, R.E.; CANDY, J.; ROSENBLUTH, M.N.

2002-01-01

OAK B202 COMPREHENSIVE GYROKINETIC SIMULATION OF TOKAMAK TURBULENCE AT FINITE RELATIVE GYRORADIUS. A continuum global gyrokinetic code GYRO has been developed to comprehensively simulate turbulent transport in actual experimental profiles and allow direct quantitative comparisons to the experimental transport flows. GYRO not only treats the now standard ion temperature gradient (ITG) mode turbulence, but also treats trapped and passing electrons with collisions and finite beta, and all in real tokamak geometry. Most importantly the code operates at finite relative gyroradius (ρ*) so as to treat the profile shear stabilization effects which break gyroBohm scaling. The code operates in a cyclic flux tube limit which allows only gyroBohm scaling and a noncyclic radial annulus with physical profile variation. The later requires an adaptive source to maintain equilibrium profiles. Simple ITG simulations demonstrate the broken gyroBohm scaling depends on the actual rotational velocity shear rates competing with mode growth rates, direct comprehensive simulations of the DIII-D ρ*-scaled L-mode experiments are presented as a quantitative test of gyrokinetics and the paradigm

DEALS magnet concept and its applcations to high density, high field tokamak systems

International Nuclear Information System (INIS)

Hsieh, S.Y.; Powell, J.; Lehner, J.; Bezler, P.; Laverick, C.; Finkelman, M.; Brown, T.; Bundy, J.

1977-01-01

The goal of the DEALS program is to develop a demountable TF magnet system concept that will reduce construction and life cycle costs, enhance the accessibility of components inside the coil system, and increase the chances for being able to use large high-field magnet systems in post TFTR reactor experiments. These experiments are projected to occur during the mid 1980's, with conceptual designs beginning in two or three years. A number of recent studies have highlighted the need for Tokamak fusion reactor systems with reasonable down time for maintenance and repair and realistic operating capacity factors, as well as the need for smaller, lower cost reactors. Two scoping studies were carried out of recent Tokamak system concepts incorporating conventionally wound coils to assess the possibilities of using demountable coils of rectangular section with an active support system and a third more intensive study using a passive support with slight movement of the joints. These studies are described briefly

Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

International Nuclear Information System (INIS)

Anderson, J.L.; Gentile, C.; Kalish, M.; Kamperschroer, J.; Kozub, T.; LaMarche, P.; Murray, H.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Sissingh, R.; Swanson, J.; Tulipano, F.; Viola, M.; Voorhees, D.; Walters, R.T.

1995-01-01

The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.)

Data acquisition and control system for the ECE imaging diagnostic on the EAST tokamak

Science.gov (United States)

Luo, C.; Lan, T.; Zhu, Y.; Xie, J.; Gao, B.; Liu, W.; Yu, C.; Milne, P. G.; Domier, C. W.; Luhmann, N. C.

2017-06-01

A 384-channel electron cyclotron emission imaging (ECEI) system is installed on the experimental advanced superconducting tokamak (EAST) and 7-gigabyte data is produced for each regular discharge of a 10-second pulse. The data acquisition and control (DAC) system for the EAST ECEI diagnostics covers the large data production and embeds the ability to report the data quality instantly after the discharge. The symmetric routing design of the timing signal distributions among the 384 channels provides a low-cost solution to the synchronization of a large number of channels. The application of the load-balance bond service largely reduces the configuration difficulty and the cost in the high-speed data transferring tasks. Benefiting from the various kinds of hardware units with dedicated functionalities, an automated and user interactive DAC work flow is achieved, including the pre-selections of the automation scheme and the observation region, 384-channel data acquisition and local caching, post-discharge imaging data quality evaluation, remote system status monitoring, and inter-discharge imaging system event handling. The system configuration in a specific physics experiment is further optimized through the associated operating software which is enhanced by the input of the tokamak operation status and the region of interest (ROI) from other diagnostics. The DAC system is based on a modularized design and scalable to the long-pulse discharges in the EAST tokamak.

Data acquisition and control system for the ECE imaging diagnostic on the EAST tokamak

International Nuclear Information System (INIS)

Luo, C.; Lan, T.; Xie, J.; Gao, B.; Liu, W.; Yu, C.; Zhu, Y.; Domier, C.W.; Luhmann, N.C.; Milne, P.G.

2017-01-01

A 384-channel electron cyclotron emission imaging (ECEI) system is installed on the experimental advanced superconducting tokamak (EAST) and 7-gigabyte data is produced for each regular discharge of a 10-second pulse. The data acquisition and control (DAC) system for the EAST ECEI diagnostics covers the large data production and embeds the ability to report the data quality instantly after the discharge. The symmetric routing design of the timing signal distributions among the 384 channels provides a low-cost solution to the synchronization of a large number of channels. The application of the load-balance bond service largely reduces the configuration difficulty and the cost in the high-speed data transferring tasks. Benefiting from the various kinds of hardware units with dedicated functionalities, an automated and user interactive DAC work flow is achieved, including the pre-selections of the automation scheme and the observation region, 384-channel data acquisition and local caching, post-discharge imaging data quality evaluation, remote system status monitoring, and inter-discharge imaging system event handling. The system configuration in a specific physics experiment is further optimized through the associated operating software which is enhanced by the input of the tokamak operation status and the region of interest (ROI) from other diagnostics. The DAC system is based on a modularized design and scalable to the long-pulse discharges in the EAST tokamak.

Numerical simulation of feedback stabilization of axisymmetric modes in tokamaks using driven halo currents

International Nuclear Information System (INIS)

Jardin, S.C.; Schmidt, J.A.

1998-01-01

The Tokamak Simulation Code (TSC) has been used to model a new method of feedback stabilization of the axisymmetric instability in tokamaks using driven halo (or scrape-off layer) currents. The method appears to be feasible for a wide range of plasma edge parameters. It may offer advantages over the more conventional method of controlling this instability when applied in a reactor environment. (author)

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

Science.gov (United States)

Chan, V. S.; Costley, A. E.; Wan, B. N.; Garofalo, A. M.; Leuer, J. A.

2015-02-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ˜ 12, Pfus ˜ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

Diagnostics for the Rijnhuizen Tokamak Project

NARCIS (Netherlands)

Donne, A. J. H.

1994-01-01

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

A self-description data framework for Tokamak control system design

Energy Technology Data Exchange (ETDEWEB)

Zhang, Ming; Zhang, Jing [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zheng, Wei, E-mail: zhengwei@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Hu, Feiran; Zhuang, Ge [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

2015-10-15

Highlights: • The SDD framework can be applied to different Tokamak devices. • We explain how configuration settings of control systems are described in SDD models, namely components and connections. • Evolving SDD models are stored in a dynamic schema database. • The SDD editor supports plug-and-play SDD models. - Abstract: A Tokamak device consists of numerous control systems, which need to be integrated. CODAC (Control, Data Access and Communication) system requires the configuration settings of these control systems to carry out the integration smoothly. SDD (Self-description data) is designed to describe the static configuration of control systems. ITER CODAC group has released an SDD software package for control system designers to manage the static configuration, but it is specific for ITER plant control systems. Following the idea of ITER SDD, we developed a flexible and scalable SDD framework to develop SDD software for J-TEXT and other sophisticated devices. The SDD framework describes the configuration settings of various control systems, including physical and logical elements and their relation information, in SDD models which are classified into Components and Connections. The framework is composed of three layers: the MongoDB database, an open-source, dynamic schema, NoSQL (Not Only SQL) database; the SDD service, which maps SDD models to MongoDB and handles the transaction and business logic; the SDD applications, which can be used to create and maintain SDD information, and generate various kinds of output using the stored SDD information.

A self-description data framework for Tokamak control system design

International Nuclear Information System (INIS)

Zhang, Ming; Zhang, Jing; Zheng, Wei; Hu, Feiran; Zhuang, Ge

2015-01-01

Highlights: • The SDD framework can be applied to different Tokamak devices. • We explain how configuration settings of control systems are described in SDD models, namely components and connections. • Evolving SDD models are stored in a dynamic schema database. • The SDD editor supports plug-and-play SDD models. - Abstract: A Tokamak device consists of numerous control systems, which need to be integrated. CODAC (Control, Data Access and Communication) system requires the configuration settings of these control systems to carry out the integration smoothly. SDD (Self-description data) is designed to describe the static configuration of control systems. ITER CODAC group has released an SDD software package for control system designers to manage the static configuration, but it is specific for ITER plant control systems. Following the idea of ITER SDD, we developed a flexible and scalable SDD framework to develop SDD software for J-TEXT and other sophisticated devices. The SDD framework describes the configuration settings of various control systems, including physical and logical elements and their relation information, in SDD models which are classified into Components and Connections. The framework is composed of three layers: the MongoDB database, an open-source, dynamic schema, NoSQL (Not Only SQL) database; the SDD service, which maps SDD models to MongoDB and handles the transaction and business logic; the SDD applications, which can be used to create and maintain SDD information, and generate various kinds of output using the stored SDD information.

Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

International Nuclear Information System (INIS)

Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

2002-01-01

A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

Transmutation of minor actinides in a spherical torus tokamak fusion reactor, FDTR

International Nuclear Information System (INIS)

Feng, K.M.; Zhang, G.S.; Deng, M.G.

2003-01-01

In this paper, a concept for the transmutation of minor actinide (MA) nuclear wastes based on a spherical torus (ST) tokamak reactor, FDTR, is put forward. A set of plasma parameters suitable for the transmutation blanket was chosen. The 2-D neutron transport code TWODANT, the 3-D Monte Carlo code MCNP/4B, the 1-D neutron transport and burn-up calculation code BISON3.0 and their associated data libraries were used to calculate the transmutation rate, the energy multiplication factor and the tritium breeding ratio of the transmutation blanket. The calculation results for the system parameters and the actinide series isotopes for different operation times are presented. The engineering feasibility of the center-post (CP) of FDTR has been investigated and the results are also given. A preliminary neutronics calculation based on an ST transmutation blanket shows that the proposed system has a high transmutation capability for MA wastes. (author)

Gas Fuelling System for SST-1Tokamak

Science.gov (United States)

Dhanani, Kalpesh; Raval, D. C.; Khan, Ziauddin; Semwal, Pratibha; George, Siju; Paravastu, Yuvakiran; Thankey, Prashant; Khan, M. S.; Pradhan, Subrata

2017-04-01

SST-1 Tokamak, the first Indian Steady-state Superconducting experimental device is at present under operation in the Institute for Plasma Research. For plasma break down & initiation, piezoelectric valve based gas feed system is implemented as a primary requirement due to its precise control, easy handling, low construction and maintenance cost and its flexibility in the selection of the working gas. Hydrogen gas feeding with piezoelectric valve is used in the SST-1 plasma experiments. The piezoelectric valves used in SST-1 are remotely driven by a PXI based platform and are calibrated before each SST-1 plasma operation with precise control. This paper will present the technical development and the results of the gas fuelling system of SST-1.

Real-time horizontal position control for Aditya-upgrade tokamak

International Nuclear Information System (INIS)

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

2015-01-01

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

Preliminary project of s Thomson scattering system for the ETE tokamak; Projeto preliminar de um sistema de espalhamento Thomson para o Tokamak ETE

Energy Technology Data Exchange (ETDEWEB)

Berni, Luiz Angelo

1997-12-31

This report presents the preliminary project of the injection and laser light block system for the Thomson (ET) scattering diagnostic to be implanted at the ETE spheric tokamak of the Instituto Nacional de Pesquisas Espaciais (INPE/LAP). Also, a scanning system for the optics of scattered light 4 refs., 26 figs.

PXIe based data acquisition and control system for ECRH systems on SST-1 and Aditya tokamak

Energy Technology Data Exchange (ETDEWEB)

Patel, Jatinkumar J., E-mail: jatin@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar (India); Shukla, B.K.; Rajanbabu, N.; Patel, H.; Dhorajiya, P.; Purohit, D. [Institute for Plasma Research, Bhat, Gandhinagar (India); Mankadiya, K. [Optimized Solutions Pvt. Ltd (India)

2016-11-15

Highlights: • Data Aquisition and control system (DAQ). • PXIe hardware–(PXI–PCI bus extension for Instrumention Express). • RHVPS–Regulated High Voltage Power supply. • SST1–Steady state superconducting tokamak. - Abstract: In Steady State Superconducting (SST-1) tokamak, various RF heating sub-systems are used for plasma heating experiments. In SST-1, Two Electron Cyclotron Resonance Heating (ECRH) systems have been installed for pre-ionization, heating and current drive experiments. The 42 GHz gyrotron based ECRH system is installed and in operation with SST-1 plasma experiments. The 82.6 GHz gyrotron delivers 200 kW CW power (1000 s) while the 42 GHz gyrotron delivers 500 kW power for 500 ms duration. Each gyrotron system consists of various auxiliary power supplies, the crowbar unit and the water cooling system. The PXIe (PCI bus extension for Instrumentation Express)bus based DAC (Data Acquisition and Control) system has been designed, developed and under implementation for safe and reliable operation of the gyrotron. The Control and Monitoring Software applications have been developed using NI LabView 2014 software with real time support on windows platform.

PXIe based data acquisition and control system for ECRH systems on SST-1 and Aditya tokamak

International Nuclear Information System (INIS)

Patel, Jatinkumar J.; Shukla, B.K.; Rajanbabu, N.; Patel, H.; Dhorajiya, P.; Purohit, D.; Mankadiya, K.

2016-01-01

Highlights: • Data Aquisition and control system (DAQ). • PXIe hardware–(PXI–PCI bus extension for Instrumention Express). • RHVPS–Regulated High Voltage Power supply. • SST1–Steady state superconducting tokamak. - Abstract: In Steady State Superconducting (SST-1) tokamak, various RF heating sub-systems are used for plasma heating experiments. In SST-1, Two Electron Cyclotron Resonance Heating (ECRH) systems have been installed for pre-ionization, heating and current drive experiments. The 42 GHz gyrotron based ECRH system is installed and in operation with SST-1 plasma experiments. The 82.6 GHz gyrotron delivers 200 kW CW power (1000 s) while the 42 GHz gyrotron delivers 500 kW power for 500 ms duration. Each gyrotron system consists of various auxiliary power supplies, the crowbar unit and the water cooling system. The PXIe (PCI bus extension for Instrumentation Express)bus based DAC (Data Acquisition and Control) system has been designed, developed and under implementation for safe and reliable operation of the gyrotron. The Control and Monitoring Software applications have been developed using NI LabView 2014 software with real time support on windows platform.

ELECTRIC POTENIAL CELLS AT THE DIVERTED TOKAMAK SEPARATRIX

International Nuclear Information System (INIS)

SCHAFFER, M.J.; PORTER, G.D.; BOEDO, J.A.; BRAY, B.D.; HSIEH, C.L.; MOYER, R.A.; ROGNLIEN, T.D.; STANGEBY, P.C.; WATKINS, J.G.

2000-01-01

OAK-B135 Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (p e ) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill (∼ 50 V) drives E x B circulation (potential cell) of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arise from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the (del)B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. They speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode

Electric potential cells at the diverted tokamak separatrix

Energy Technology Data Exchange (ETDEWEB)

Schaffer, Michael J.; Bray, Bruce D.; Hsieh, Chung-Lih [General Atomics, San Diego, California (United States); Porter, Gary D.; Rognlien, Thomas D. [Lawrence Livermore National Laboratory, Livermore, California (United States); Boedo, Jose A.; Moyer, Richard A. [University of California, San Diego, California (United States); Stangeby, Peter C. [Univ. of Toronto, Toronto (Canada); Watkins, Jonathan G. [Sandia National Laboratories, Albuquerque, New Mexico (United States)

2001-07-01

Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (p{sub e}) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill ({approx}100 V) drives ExB circulation ('potential cell') of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arises from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the {nabla}B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. We speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode. (author)

Electric potential cells at the diverted tokamak separatrix

International Nuclear Information System (INIS)

Schaffer, Michael J.; Bray, Bruce D.; Hsieh, Chung-Lih; Porter, Gary D.; Rognlien, Thomas D.; Boedo, Jose A.; Moyer, Richard A.; Stangeby, Peter C.; Watkins, Jonathan G.

2001-01-01

Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (p e ) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill (∼100 V) drives ExB circulation ('potential cell') of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arises from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the ∇B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. We speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode. (author)

Study of the linear and non-linear coupling of the LH wave to the tokamak plasmas

International Nuclear Information System (INIS)

Preynas, M.

2012-10-01

In order to achieve long pulse operation with a tokamak, additional heating and current drive systems are necessary. High frequency antennas, which deliver several megawatts of power to the plasma, are currently used in several tokamaks. Moreover, a good control of the coupling of the wave launched by the antenna to the edge plasma is required to optimize the efficiency of heating and current drive LH systems. However, non-linear effects which depend on the level of injected power in the plasma strongly damage the coupling of the LH wave at particular edge parameters (density and temperature profiles). Results presented in the manuscript deal with the study of the linear and non-linear coupling of the LH wave to the plasma. In the framework of the commissioning of the Passive Active Multijunction antenna in 2009 on the Tore Supra tokamak aiming at validating the LH system suggested for ITER, the characterisation of its coupling properties was realized from low power experiments. The experimental results, which are compared with the linear coupling code ALOHA, have validated the theoretical predictions of good coupling at edge plasma density around the cut-off density. Besides, the ponderomotive effect is clearly identified as responsible for the deterioration in the coupling of the wave, which is measured under particular edge plasma conditions. A theoretical model combining the coupling of the LH wave with the ponderomotive force is suggested to explain the experimental observations. Thus, a new full wave code (named PICCOLO-2D) was developed and results from simulations validate the working hypothesis of the contribution of the ponderomotive effect in the non-linear observations of LHCD coupling on Tore Supra. (author)

The control system position to the electric probe in the Tokamak Novillo

International Nuclear Information System (INIS)

Sanchez Garcia, A.M.

1993-01-01

The electric probe are used to determine the parameters of electronic temperatures, the electron density and the plasma potential in Tokamak machines. On this machines the electric probes are used only in the plasma edge due to the intensive flow of high energy particles. This is the region in which the plasma density and temperature are relatively low. It is showed, in this work, the design and construction of an electro mechanic system which is used to control the position of the probe into the discharge chamber. This system is called T he control system position to the electric probe in the tokamak Novillo . This controller is a minimum system that is in charge , by a programming, to rule a step motor by a logic sequence commutation. This is done with the purpose of slide the probe in a radial way with a milli metric precision into the discharge chamber. To this purpose it is used a step motor, due it is principal characteristic is the control of the end element position without a feedback needing of the wrong signal. The system function consist on reading, through a board, the corresponding data to the position where it is wanted to place the probe, it also displays by a numeric indicator the position in which the probe is located (in an interval from 0 to 100 mm), and provide the logic sequence commutation for the step motor. The minimum system is constituted by the micro controller 8748-8 that gives with all precision the control of the electric probe position in the Tokamak Novillo, by programming, associated circuits, amplification unit bi phase unipolar and switching power (they supply the power to the control circuit and to the step motor too), avoiding the destruction of the electric probe. (Author). 17 refs, 29 figs

Analysis of toroidal rotation data for the DIII-D tokamak

International Nuclear Information System (INIS)

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

1989-01-01

Both poloidal and toroidal rotation are observed during routine neutral beam heating operation of the DIII-D tokamak. Poloidal rotation results and the empirical techniques used to measure toroidal and poloidal rotation speeds are described by Groebner. Here we concentrate on the analysis of recent measurements of toroidal rotation made during diverted, H-mode operation of the DIII-D tokamak during co- and counter-neutral beam injection of hydrogen into deuterium plasmas. Our results are based on numerical inversions using the transport code ONETWO, modified to account for the radial diffusion of toroidal angular momentum. 13 refs., 4 figs

Axisymmetric stability of vertically asymmetric tokamaks at large beta poloidal

International Nuclear Information System (INIS)

Yamazaki, K.; Fishman, H.; Okabayashi, M.; Todd, A.M.M.

1981-09-01

The stability of high-β vertically asymmetric tokamak equilibria to rigid displacements is investigated analytically. It is found that vertical stability at large beta poloidal is mainly determined by a coupling between the shape of the plasma surface and the Shafranov shift of the magnetic axis. To the lowest order, symmetric components of the plasma surface shape are found to be the critical destabilizing elements. Asymmetric components have little effect. The inclusion of higher order terms in the high β tokamak expansion leads to further destabilization. Qualitative agreement between these analytic results and numerical stability calculations using the PEST code is demonstrated

Three-dimensional tokamak equilibria and stellarators with two-dimensional magnetic symmetry

International Nuclear Information System (INIS)

Garabedian, P.R.

1997-01-01

Three-dimensional computer codes have been developed to simulate equilibrium, stability and transport in tokamaks and stellarators. Bifurcated solutions of the tokamak problem suggest that three-dimensional effects may be more important than has generally been thought. Extensive calculations have led to the discovery of a stellarator configuration with just two field periods and with aspect ratio 3.2 that has a magnetic field spectrum B mn with toroidal symmetry. Numerical studies of equilibrium, stability and transport for this new device, called the Modular Helias-like Heliac 2 (MHH2), will be presented. (author)

Design parameters of Tokamak-7 system

International Nuclear Information System (INIS)

Ivanov, D.P.; Keilin, V.E.; Klimenko, E.Yu.; Strelkov, V.S.

Superconducting windings for the main magnetic field of Tokamak-7 are discussed. The parameters of this facility are based on the use of commercially available superconducting materials for fields up to 80 kOe. Experimental parameters are described. (U.S.)

Digital control of plasma position in Damavand tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2002-03-01

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

A distributed control system for the lower-hybrid current drive system on the Tokamak de Varennes

International Nuclear Information System (INIS)

Bagdoo, J.; Guay, J.M.; Chaudron, G.A.; Decoste, R.; Demers, Y.; Hubbard, A.

1990-01-01

An rf current drive system with an output power of 1 MW at 3.7 GHz is under development for the Tokamak de Varennes. The control system is based on an Ethernet local-area network of programmable logic controllers as front end, personal computers as consoles, and CAMAC-based DSP processors. The DSP processors ensure the PID control of the phase and rf power of each klystron, and the fast protection of high-power rf hardware, all within a 40 μs loop. Slower control and protection, event sequencing and the run-time database are provided by the programmable logic controllers, which communicate, via the LAN, with the consoles. The latter run a commercial process-control console software. The LAN protocol respects the first four layers of the ISO/OSI 802.3 standard. Synchronization with the tokamak control system is provided by commercially available CAMAC timing modules which trigger shot-related events and reference waveform generators. A detailed description of each subsystem and a performance evaluation of the system will be presented. (orig.)

A distributed control system for the lower-hybrid current drive system on the Tokamak de Varennes

Science.gov (United States)

Bagdoo, J.; Guay, J. M.; Chaudron, G.-A.; Decoste, R.; Demers, Y.; Hubbard, A.

1990-08-01

An rf current drive system with an output power of 1 MW at 3.7 GHz is under development for the Tokamak de Varennes. The control system is based on an Ethernet local-area network of programmable logic controllers as front end, personal computers as consoles, and CAMAC-based DSP processors. The DSP processors ensure the PID control of the phase and rf power of each klystron, and the fast protection of high-power rf hardware, all within a 40 μs loop. Slower control and protection, event sequencing and the run-time database are provided by the programmable logic controllers, which communicate, via the LAN, with the consoles. The latter run a commercial process-control console software. The LAN protocol respects the first four layers of the ISO/OSI 802.3 standard. Synchronization with the tokamak control system is provided by commercially available CAMAC timing modules which trigger shot-related events and reference waveform generators. A detailed description of each subsystem and a performance evaluation of the system will be presented.

Limiter/vacuum system for plasma impurity control and exhaust in tokamaks

International Nuclear Information System (INIS)

Abdou, M.; Brooks, J.; Mattas, R.

1980-01-01

A detailed design of a limiter/vacuum system for plasma impurity control and exhaust has been developed for the STARFIRE tokamak power plant. It is shown that the limiter/vacuum concept is a very attractive option for power reactors. It is relatively simple and inexpensive and deserves serious experimental verification

StructUre and test results of the Tokamak-7 device cryogenic system

International Nuclear Information System (INIS)

Babaev, I.V.; VolobUev, A.N.; Zhul'kin, V.F.

1982-01-01

A cryogenic system (CS) of the Tokamak-7 (T-7) installation with the longitudinal field superconducting magnetic system (SMS) is described. The CS is designed for cool-down, cryostatic cooling and heating of the T-7 cryogenic objects and consists of a helium system (HS) and a nitrogen cryogenic system (NCS). The HS consists of:a a heliUm delivery system intended for distributing and controlling the helium flows in the SMS; cryogenic helium units; a 1.25 m 3 volume for storing liquid helium; a compressor compartment using piston compressors at the 3 MPa operating pressure and 140 g/s total capacity; gaseous helium storages (3600 m 3 under normal conditions); helium cleaning and drying systems; a gas holder of 20 m 3 operating volume; cryogenic pipelines and pipe fittings. The NCS operates on delivered nitrogen and includes a 120 m 3 liquid nitrogen storage, evaporators and electric heaters producing up to 230 g/s of gaseous nitrogen at 300 K, a separator, cryogenic pipelines and fittings. It is found that the CS has the necessary cold production reserve, ensures reliable operation of the Tokamak-7 device and permits to carry out practically continuous plasma experiments

Simulation models for tokamak plasmas

International Nuclear Information System (INIS)

Dimits, A.M.; Cohen, B.I.

1992-01-01

Two developments in the nonlinear simulation of tokamak plasmas are described: (A) Simulation algorithms that use quasiballooning coordinates have been implemented in a 3D fluid code and a 3D partially linearized (Δf) particle code. In quasiballooning coordinates, one of the coordinate directions is closely aligned with that of the magnetic field, allowing both optimal use of the grid resolution for structures highly elongated along the magnetic field as well as implementation of the correct periodicity conditions with no discontinuities in the toroidal direction. (B) Progress on the implementation of a likeparticle collision operator suitable for use in partially linearized particle codes is reported. The binary collision approach is shown to be unusable for this purpose. The algorithm under development is a complete version of the test-particle plus source-field approach that was suggested and partially implemented by Xu and Rosenbluth

Comparison between stellarator and tokamak divertor transport

International Nuclear Information System (INIS)

Feng, Y.; Lunt, T.; Kobayashi, M.; Reiter, D.

2010-11-01

The paper compares the essential divertor transport features of the poloidal divertor, which is well-developed for tokamaks, and the non-axisymmetric divertors currently investigated on helical devices. It aims at surveying the fundamental similarities and differences in divertor concept and geometry, and their consequences for how the divertor functions. In particular, the importance of various transport terms governing axisymmetric and helical scrape-off-layers (SOLs) is examined, with special attention being paid to energy, momentum and impurity transport. Tokamak and stellarator SOLs are compared by identifying key geometric parameters through which the governing physics can be illustrated by simple models and estimates. More quantitative assessments rely nevertheless on the modeling using EMC3-EIRENE code. Most of the theoretical results are discussed in conjunction with experimental observations. (author)

Constrained ripple optimization of Tokamak bundle divertors

International Nuclear Information System (INIS)

Hively, L.M.; Rome, J.A.; Lynch, V.E.; Lyon, J.F.; Fowler, R.H.; Peng, Y-K.M.; Dory, R.A.

1983-02-01

Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ω B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple ( 0 ) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded

Power supply system on HT-7 tokamak for diagnostic neutral beam based on PLC

International Nuclear Information System (INIS)

Zhang Jian; Liu Baohua; Ding Tonghai; Du Shaowu

2006-01-01

A power supply system for diagnostic neutral beam on the HT-7 Tokamak was developed. Its logic control system based on S7-300 PLC was described. The experimental results show that the system is easy to operate and its performance is reliable. (authors)

DIII-D Integrated plasma control solutions for ITER and next-generation tokamaks

International Nuclear Information System (INIS)

Humphreys, D.A.; Ferron, J.R.; Hyatt, A.W.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; In, Y.

2008-01-01

Plasma control design approaches and solutions developed at DIII-D to address its control-intensive advanced tokamak (AT) mission are applicable to many problems facing ITER and other next-generation devices. A systematic approach to algorithm design, termed 'integrated plasma control,' enables new tokamak controllers to be applied operationally with minimal machine time required for tuning. Such high confidence plasma control algorithms are designed using relatively simple ('control-level') models validated against experimental response data and are verified in simulation prior to operational use. A key element of DIII-D integrated plasma control, also required in the ITER baseline control approach, is the ability to verify both controller performance and implementation by running simulations that connect directly to the actual plasma control system (PCS) that is used to operate the tokamak itself. The DIII-D PCS comprises a powerful and flexible C-based realtime code and programming infrastructure, as well as an arbitrarily scalable hardware and realtime network architecture. This software infrastructure provides a general platform for implementation and verification of realtime algorithms with arbitrary complexity, limited only by speed of execution requirements. We present a complete suite of tools (known collectively as TokSys) supporting the integrated plasma control design process, along with recent examples of control algorithms designed for the DIII-D PCS. The use of validated physics-based models and a systematic model-based design and verification process enables these control solutions to be directly applied to ITER and other next-generation tokamaks

Numerical reconstruction of azimuthal balance in the Novillo Tokamak

International Nuclear Information System (INIS)

Barocio, S.; Avila, R.; Chavez A, E.; Melendez, L.

2001-01-01

The calculation of the magneto-hydrodynamic of free bounds axially symmetric allows to predict the field distribution and currents starting from existing forces among them. So, for the case of the Novillo Tokamak a code based in an iterative strategy of finite differences is presented where the limiter creates an empty region outside of the plasma column. It is showed how the balance possess a strong dependence in the profile of current density and some examples of the code application. (Author)

Gyrokinetic global analysis of ion temperature gradient driven mode in reversed shear tokamaks

International Nuclear Information System (INIS)

Idomura, Y.; Tokuda, S.; Kishimoto, Y.

2003-01-01

A new toroidal gyrokinetic particle code has been developed to study the ion temperature gradient driven (ITG) turbulence in reactor relevant tokamak parameters. We use a new method based on a canonical Maxwellian distribution F CM (P φ , ε, μ), which is defined by three constants of motion in the axisymmetric toroidal system, the canonical angular momentum P φ , the energy ε, and the magnetic moment μ. A quasi-ballooning representation enables linear and nonlinear high-m,n global calculations with a good numerical convergence. Conservation properties are improved by using the optimized loading method. From comprehensive linear global analyses over a wide range of an unstable toroidal mode number spectrum (n=0∼100) in large tokamak parameters (a/ρ ti =320∼460), properties of the ITG modes in reversed shear tokamaks are discussed. In the nonlinear simulation, it is found that a new method based on F CM can simulate a zonal flow damping correctly, and spurious zonal flow oscillations, which are observed in a conventional method based on a local Maxwellian distribution F LM (ψ, ε, μ), do not appear in the nonlinear regime. (author)

Final technical report for DE-SC00012633 AToM (Advanced Tokamak Modeling)

Energy Technology Data Exchange (ETDEWEB)

Holland, Christopher [Univ. of California, San Diego, CA (United States); Orlov, Dmitri [Univ. of California, San Diego, CA (United States); Izzo, Valerie [Univ. of California, San Diego, CA (United States)

2018-02-05

This final report for the AToM project documents contributions from University of California, San Diego researchers over the period of 9/1/2014 – 8/31/2017. The primary focus of these efforts was on performing validation studies of core tokamak transport models using the OMFIT framework, including development of OMFIT workflow scripts. Additional work was performed to develop tools for use of the nonlinear magnetohydrodynamics code NIMROD in OMFIT, and its use in the study of runaway electron dynamics in tokamak disruptions.

HL-1 tokamak data acquisition system and its initial application in the physical experiment

International Nuclear Information System (INIS)

Deng Huichen; Fu Bo; Dong Jiafu

1989-11-01

A HL-1 tokamak data acquisition system has been developed and has been used in the physical experiment. The hardware and software configuration of the system, as well as the typical acquired data in the HL-1 experiment are introduced

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

International Nuclear Information System (INIS)

Chan, V.S.; Garofalo, A.M.; Leuer, J.A.; Costley, A.E.; Wan, B.N.

2015-01-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher B T and I p can result in a high gain Q fus  ∼ 12, P fus  ∼ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target. (paper)

Texas Experimental Tokamak

International Nuclear Information System (INIS)

Wootton, A.J.

1990-04-01

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

Microwave polarimetry system in the CDX-U tokamak

International Nuclear Information System (INIS)

Hwang, Y.S.; Fredriksen, A.; Qin, H.; Forest, C.B.; Ono, M.

1995-01-01

An existing microwave interferometer system is modified to add the capability of polarimetry in the CDX-U tokamak. Though this interferometer system can scan vertically and radially, only the vertical view channel is modified to accomodate Faraday rotation measurements, with its radial scanning capability preserved. For our relatively long microwave wavelength, the signal amplitude variation due to refraction is more important than effects due to vibration. An amplitude independent design of Faraday rotation diagnostics has been developed. By using a linearly polarized beam as input and putting a rotating polarizer in the beam after the plasma, birefringency effects are minimized. A digital phase detection technique has been developed for better resolution of the Faraday rotation angle

Equilibrium reconstruction in the TCA/Br tokamak

International Nuclear Information System (INIS)

Sa, Wanderley Pires de

1996-01-01

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

Sustained high βN plasmas on EAST tokamak

Science.gov (United States)

Gao, Xiang; the EAST team

2018-05-01

Sustained high normalized beta (βN ∼ 1.9) plasmas with an ITER-like tungsten divertor have been achieved on EAST tokamak recently. The high power NBI heating system of 4.8 MW and the 4.6 GHz lower hybrid wave of 1 MW were developed and applied to produce edge and internal transport barriers in high βN discharges. The central flat q profile with q (ρ) ∼ 1 at ρ safety factor q95 = 4.7 is identified by the multi-channel far-infrared laser polarimeter and the EFIT code. The fraction of non-inductive current is about 40%. The relation between fishbone activity and ITB formation is observed and discussed.

Major disruption process in tokamak

International Nuclear Information System (INIS)

Kurita, Gen-ichi; Azumi, Masafumi; Tuda, Takashi; Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji; Itoh, Kimitaka; Takeda, Tatsuoki

1981-11-01

The major disruption in a cylindrical tokamak is investigated by using the multi-helicity code, and the destabilization of the 3/2 mode by the mode coupling with the 2/1 mode is confirmed. The evolution of the magnetic field topology caused by the major disruption is studied in detail. The effect of the internal disruption on the 2/1 magnetic island width is also studied. The 2/1 magnetic island is not enhanced by the flattening of the q-profile due to the internal disruption. (author)

Tokamak power systems studies, FY 1986: A second stability power reactor

International Nuclear Information System (INIS)

Ehst, D.; Baker, C.; Billone, M.

1987-03-01

This report presents the results of the work at Argonne National Laboratory (ANL) during FY-1986 on the Tokamak Power Systems Study (TPSS). The purpose of the TPSS is to explore and develop ideas that would lead to improvements in the tokamak as a power reactor concept. The work at ANL concentrated on plasma engineering, impurity control, and the blanket/first wall/shield system. The work in FY-1986 extended these studies and focused them on a reference design point. The key features of the design point include: second stability regime with higher β and larger aspect ratio, steady-state operation with fast wave current drive, impurity control via a self-pumped slot limiter, a self-cooled liquid lithium, vanadium alloy blanket with simplified poloidal flow, and reduced reactor building volume with vertical lift maintenance. Sufficient work was carried out to report a preliminary cost estimate. In addition, reactor implications of steady-state operation in the first stability regime were also studied. 174 refs., 124 figs., 65 tabs

A fully implicit Newton-Krylov-Schwarz method for tokamak magnetohydrodynamics: Jacobian construction and preconditioner formulation

KAUST Repository

Reynolds, Daniel R.

2012-01-01

Single-fluid resistive magnetohydrodynamics (MHD) is a fluid description of fusion plasmas which is often used to investigate macroscopic instabilities in tokamaks. In MHD modeling of tokamaks, it is often desirable to compute MHD phenomena to resistive time scales or a combination of resistive-Alfvén time scales, which can render explicit time stepping schemes computationally expensive. We present recent advancements in the development of preconditioners for fully nonlinearly implicit simulations of single-fluid resistive tokamak MHD. Our work focuses on simulations using a structured mesh mapped into a toroidal geometry with a shaped poloidal cross-section, and a finite-volume spatial discretization of the partial differential equation model. We discretize the temporal dimension using a fully implicit or the backwards differentiation formula method, and solve the resulting nonlinear algebraic system using a standard inexact Newton-Krylov approach, provided by the sundials library. The focus of this paper is on the construction and performance of various preconditioning approaches for accelerating the convergence of the iterative solver algorithms. Effective preconditioners require information about the Jacobian entries; however, analytical formulae for these Jacobian entries may be prohibitive to derive/implement without error. We therefore compute these entries using automatic differentiation with OpenAD. We then investigate a variety of preconditioning formulations inspired by standard solution approaches in modern MHD codes, in order to investigate their utility in a preconditioning context. We first describe the code modifications necessary for the use of the OpenAD tool and sundials solver library. We conclude with numerical results for each of our preconditioning approaches in the context of pellet-injection fueling of tokamak plasmas. Of these, our optimal approach results in a speedup of a factor of 3 compared with non-preconditioned implicit tests, with

Plasma position control in TCABR Tokamak

International Nuclear Information System (INIS)

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

1998-01-01

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

Design of the RF system for Alfven wave heating and current drive in a TCA/BR tokamak

International Nuclear Information System (INIS)

Ruchko, L.; Andrade, M.L.; Ozono, E.; Galvao, R.M.O.; Degaspari, F.T.; Nascimento, I.C.

1995-01-01

The advanced RF system for Alfven wave plasma heating and current drive in TCA/BR tokamak is presented. The antenna system is capable of exciting the standing and travelling wave M = -1,N = 1,N =-4,-6 with single helicity and thus provides the possibility to improve Alfven wave plasma heating efficiency in TCA/BR tokamak and to increase input power level up to P ≅ 1 MW, without the uncontrolled density rise which was encountered in previous TCA (Switzerland) experiments. (author). 4 refs., 3 figs

Computational images of internal-transport-barrier oscillations in tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Bizarro, J.P S. [Inst Super Tecn, Ctr Fusao Nucl, EURATOM Assoc, P-1049001 Lisbon (Portugal); Litaudon, X.L. [CEA Cadarache, Dept Rech Fus Controlee, EURATOM Assoc, F-13108 St Paul Les Durance (France); Tala, T.J.J. [Assoc Euratom Tekes, FIN-02044 Espoo (Finland); JET EFDA Contributors [Culham Sci Ctr, Abingdon OX14 3DB, Oxon (United Kingdom)

2008-07-01

A well-known benchmarked code, where a Bohm-gyro-Bohm transport model is complemented with an empirical scaling for the dynamics of internal transport barriers (ITBs), is used to model the ITB oscillations that are often seen in advanced tokamak scenarios with a dominant fraction of bootstrap current. (authors)

Evolution of the DINA-CH tokamak full discharge simulator

International Nuclear Information System (INIS)

Lister, J.B.; Dokouka, V.N.; Khayrutdinov, R.R.; Lukash, V.E.; Duval, B.P.; Moret, J.-M.; Artaud, J.-F.; Baziuk, V.; Cavinato, M.

2005-01-01

This paper summarises the approach taken to develop an open architecture full tokamak discharge simulator - DINA-CH - based on the DINA code and implemented under graphical programming control using Matlab-SIMULINK. The evolution path and present status are presented, with applications to ITER and TCV. The future evolution combining DINA-CH with Cronos, is discussed

Control, pressure perturbations, displacements, and disruptions in highly elongated tokamak plasmas

International Nuclear Information System (INIS)

Marcus, F.B.; Hofmann, F.; Tonetti, G.; Jardin, S.C.; Noll, P.

1989-06-01

The control and evolution of highly elongated tokamak plasmas with large growth rates are simulated with the axisymmetric, resistive MHD code TSC in the geometry of the TCV tokamak. Pressure perturbations such as sawteeth and externally programmed displacements create initial velocity perturbations which may be stabilized by low power, rapid response coils inside the passively stabilizing vacuum vessel, together with slower shaping coils outside the vessel. Vertical disruption induced voltages and forces on the rapid coils and vessel are investigated, and a model is proposed for an additional vertical force due to poloidal currents. (author) 6 figs., 1 tab., 26 refs

Spectroscopic system for impurity measurements in the TJ-1 Tokamak of JEN; Un sistema espectroscopico para medidas de impurezas en el Tokamak TJ-1 de la JEN

Energy Technology Data Exchange (ETDEWEB)

Navas, G; Zurro, B

1982-07-01

we describe a spectroscopic system with spatial resolution capability that has been configured for plasma diagnostic in the TJ-1 Tokamak of JEN. The experimental system, based on a one meter monochromator, has been absolutely calibrated using a tungsten-halogen lamp. The calibration procedures and the absolute spectral sensitivity are presented as well as its dependence with the polarization. A simplified spectroscopic model of the radiation emitted by the intrinsic plasma impurities (C, 0, . . . ) has been developed. A one dimensional model of the temporal evolution of various ionization stages in coronal equilibrium is used to predict the electron temperature and impurity concentration. This model has been applied to experimental data from several Tokamaks. (Author) 23 refs.

Neutronics design for a spherical tokamak fusion-transmutation reactor

International Nuclear Information System (INIS)

Deng Meigen; Feng Kaiming; Yang Bangchao

2002-01-01

Based on studies of the spherical tokamak fusion reactors, a concept of fusion-transmutation reactor is put forward. By using the one-dimension transport and burn-up code BISON3.0 to process optimized design, a set of plasma parameters and blanket configuration suitable for the transmutation of MA (Minor Actinides) nuclear waste is selected. Based on the one-dimension calculation, two-dimension calculation has been carried out by using two-dimension neutronics code TWODANT. Combined with the neutron flux given by TWODANT calculation, burn-up calculation has been processed by using the one-dimension radioactivity calculation code FDKR and some useful and reasonable results are obtained

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

Electron Bernstein wave current drive in the start-up phase of a tokamak discharge

International Nuclear Information System (INIS)

Montes, A.; Ludwig, G.O.

1986-04-01

Current drive by electron Bernstein waves in the start-up phase of tokamak discharges is studied. A general analytical expression is derived for the figure of merit J/Pd associated with these waves. This is coupled with a ray tracing code, allowing the calculation of the total current generated per unit of incident power in realistic tokamak conditions. The resuts show that the electron Bernstein waves can drive substantial currents even at very low electron temperatures. (Author) [pt

Internal transport barrier formation and pellet injection simulation in helical and tokamak reactors

International Nuclear Information System (INIS)

Higashiyama, You; Yamazaki, Kozo; Arimoto, Hideki; Garcia, Jeronimo

2008-01-01

In the future fusion reactor, plasma density peaking is important for increase in the fusion power gain and for achievement of confinement improvement mode. Density control and internal transport barrier (ITB) formation due to pellet injection have been simulated in tokamak and helical reactors using the toroidal transport linkage code TOTAL. First, pellet injection simulation is carried out, including the neutral gas shielding model and the mass relocation model in the TOTAL code, and the effectiveness of high-field side (HFS) pellet injection is clarified. Second, ITB simulation with pellet injection is carried out with the confinement improvement model based on the E x B shear effects, and it is found that deep pellet penetration is helpful for ITB formation as well as plasma core fuelling in the reversed-shear tokamak and helical reactors. (author)

Automated Fault Detection for DIII-D Tokamak Experiments

International Nuclear Information System (INIS)

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

1999-01-01

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

The superconducting magnet system for the Tokamak Physics Experiment

International Nuclear Information System (INIS)

Lang, D.D.; Bulmer, R.J.; Chaplin, M.R.; O'Connor, T.G.; Slack, D.S.; Wong, R.L.; Zbasnik, J.P.; Schultz, J.H.; Diatchenko, N.; Montgomery, D.B.

1994-01-01

The superconducting magnet system for the Tokamak Physics eXperiment (TPX) will be the first all superconducting magnet system for a Tokamak, where the poloidal field coils, in addition to the toroidal field coils are superconducting. The magnet system is designed to operate in a steady state mode, and to initiate the plasma discharge ohmically. The toroidal field system provides a peak field of 4.0 Tesla on the plasma axis at a plasma major radius of 2.25 m. The peak field on the niobium 3-tin, cable-in-conduit (CIC) conductor is 8.4 Tesla for the 16 toroidal field coils. The toroidal field coils must absorb approximately 5 kW due to nuclear heating, eddy currents, and other sources. The poloidal field system provides a total of 18 volt seconds to initiate the plasma and drive a plasma current up to 2 MA. The poloidal field system consists of 14 individual coils which are arranged symmetrically above and below the horizontal mid plane. Four pairs of coils make up the central solenoid, and three pairs of poloidal ring coils complete the system. The poloidal field coils all use a cable-in-conduit conductor, using either niobium 3-tin (Nb 3 Sn) or niobium titanium (NbTi) superconducting strands depending on the operating conditions for that coil. All of the coils are cooled by flowing supercritical helium, with inlet and outlet connections made on each double pancake. The superconducting magnet system has gone through a conceptual design review, and is in preliminary design started by the LLNL/MIT/PPPL collaboration. A number of changes have been made in the design since the conceptual design review, and are described in this paper. The majority of the design and all fabrication of the superconducting magnet system will be ,accomplished by industry, which will shortly be taking over the preliminary design. The magnet system is expected to be completed in early 2000

The scientific program of the Tokamak de Varennes

International Nuclear Information System (INIS)

Daughney, C.C.

1989-01-01

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

Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Hartmann, Tobias

2013-07-03

The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

International Nuclear Information System (INIS)

Hartmann, Tobias

2013-01-01

The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

Tokamak engineering mechanics

International Nuclear Information System (INIS)

Song, Yuntao; Wu, Weiyue; Du, Shijun

2014-01-01

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

Neutral beam injection system design for KSTAR tokamak

Energy Technology Data Exchange (ETDEWEB)

Choi, B.H.; Lee, K.W.; Chung, K.S.; Oh, B.H.; Cho, Y.S.; Bae, Y.D.; Han, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-06-01

The NBI system for KSTAR (Korean Superconducting Tokamak Advanced Research) has been designed based on conventional positive ion beam technology. One beam line consists of three ion sources, three neutralizers, one bending magnet, and one drift tube. This system will deliver 8 MW deuterium beam to KSTAR plasma in normal operation to support the advanced experiments on heating, current drive and profile control. The key technical issues in this design were high power ion source(120 kV, 65 A), long pulse operation (300 seconds; world record is 30 sec), and beam rotation from vertical to horizontal direction. The suggested important R and D points on ion source and beam line components are also included. (author). 7 refs., 27 figs., 1 tab.

Simulation of saturated tearing modes in tokamaks

International Nuclear Information System (INIS)

Nguyen, Canh N.; Bateman, Glenn; Kritz, Arnold H.

2004-01-01

A quasi-linear model, which includes the effect of the neoclassical bootstrap current, is developed for saturated tearing modes in order to compute magnetic island widths in axisymmetric toroidal plasmas with arbitrary aspect ratio and cross-sectional shape. The model is tested in a simple stand-alone code and is implemented in the BALDUR [C. E. Singer et al., Comput. Phys. Commun. 49, 275 (1982)] predictive modeling code. It is found that the widths of tearing mode islands increase with decreasing aspect ratio and with increasing elongation. Also, the island widths increase when the gradient of the current density increases at the edge of the islands and when the current density inside the islands is suppressed, such as the suppression caused by the near absence of the bootstrap current within the islands. In simulations of tokamak discharges, it is found that tearing mode island widths oscillate in time in response to periodic sawtooth crashes. The local enhancements in the transport produced by magnetic islands have a noticeable effect on global plasma confinement in simulations of low aspect ratio, high beta tokamaks, where saturated tearing mode islands can occur with widths that are greater than 15% of the plasma minor radius

Tokamak configuration analysis with the method of toroidal multipoles

International Nuclear Information System (INIS)

Micozzi, P.; Alladio, F.; Crisanti, F.; Marinucci, M.; Tanga, A.

1989-01-01

In the study of tokamak machines able to sustain plasmas of thermonuclear interest (JIT, IGNITOR, NET, CIT, ET), there is a strong quest for engineering optimization of the circuital components close to the plasma. We have developed a semianalytical axisymmetric MHD equilibrium code based on the technique of the poloidal ψ flux function expansion in toroidal harmonic series. This code is able to optimize the necessary currents in the poloidal circuits in order to sustain a plasma of fixed shape (also x-point configuration), toroidal current and poloidal β. (author) 4 refs., 4 figs

The implementation of a toroidal limiter model into the gyrokinetic code ELMFIRE

Energy Technology Data Exchange (ETDEWEB)

Leerink, S.; Janhunen, S.J.; Kiviniemi, T.P.; Nora, M. [Euratom-Tekes Association, Helsinki University of Technology (Finland); Heikkinen, J.A. [Euratom-Tekes Association, VTT, P.O. Box 1000, FI-02044 VTT (Finland); Ogando, F. [Universidad Nacional de Educacion a Distancia, Madrid (Spain)

2008-03-15

The ELMFIRE full nonlinear gyrokinetic simulation code has been developed for calculations of plasma evolution and dynamics of turbulence in tokamak geometry. The code is applicable for calculations of strong perturbations in particle distribution function, rapid transients and steep gradients in plasma. Benchmarking against experimental reflectometry data from the FT2 tokamak is being discussed and in this paper a model for comparison and studying poloidal velocity is presented. To make the ELMFIRE code suitable for scrape-off layer simulations a simplified toroidal limiter model has been implemented. The model is be discussed and first results are presented. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Tokamak concept innovations

International Nuclear Information System (INIS)

1986-04-01

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

Design of an O-mode frequency modulated reflectometry system for the measurement of Alborz Tokamak plasma density profile

Energy Technology Data Exchange (ETDEWEB)

Koohestani, Saeideh [Department of Energy Engineering and physics, Amirkabir University of Technology, Tehran, 15875-4413, Islamic Republic of Iran (Iran, Islamic Republic of); Amrollahi, Reza, E-mail: amrollahi@aut.ac.ir [Department of Energy Engineering and physics, Amirkabir University of Technology, Tehran, 15875-4413, Islamic Republic of Iran (Iran, Islamic Republic of); Moradi, Gholamreza [Department of Electrical Engineering, Amirkabir University of Technology, Tehran, 15875-4413, Islamic Republic of Iran (Iran, Islamic Republic of)

2016-12-15

Reflectometry is a common method for plasma diagnostic, in which microwaves are launched into the plasma and reflected at the critical surfaces. Comparing the reflected microwave signals with the launched waves would give rise to the plasma density profiles. In the present study, an ordinary mode (O-mode) frequency modulation (FM) reflectometry system has been designed for the electron density profile measurement of the Alborz Tokamak plasma. This system has been considered to operate at K-band (18–26.5 GHz) frequency range and scan the frequency band between 18 to 26 GHz in 40 μS. The density profile from major radius r = 47.9–51.55 cm can be measured in Alborz Tokamak plasma. Based on the Alborz Tokamak operational conditions, the characteristic frequencies, and some dimensional limitations, all parts of reflectometer have been designed so that an appropriate efficiency with minimum attenuation, especially in transmitting/receiving system would be achieved. A dual antenna and an oversized waveguide of X-band (8–12 GHz) for transmitting and receiving purposes and a balanced detector for absolute phase determination have been utilized. The details of the Alborz Tokamak FM reflectometry components focusing on the antenna and waveguide design and mounting are described in this paper. Additionally, the procedure of plasma profile reconstruction using the system output signal is discussed. This system uses signal phase shift to determine the position of the cutoff layer.

Modification of PRETOR Code to Be Applied to Transport Simulation in Stellarators

International Nuclear Information System (INIS)

Fontanet, J.; Castejon, F.; Dies, J.; Fontdecaba, J.; Alejaldre, C.

2001-01-01

The 1.5 D transport code PRETOR, that has been previously used to simulate tokamak plasmas, has been modified to perform transport analysis in stellarator geometry. The main modifications that have been introduced in the code are related with the magnetic equilibrium and with the modelling of energy and particle transport. Therefore, PRETOR- Stellarator version has been achieved and the code is suitable to perform simulations on stellarator plasmas. As an example, PRETOR- Stellarator has been used in the transport analysis of several Heliac Flexible TJ-II shots, and the results are compared with those obtained using PROCTR code. These results are also compared with the obtained using the tokamak version of PRETOR to show the importance of the introduced changes. (Author) 18 refs

BALDUR: a one-dimensional plasma transport code

International Nuclear Information System (INIS)

Singer, C.E.; Post, D.E.; Mikkelsen, D.R.

1986-07-01

The purpose of BALDUR is to calculate the evolution of plasma parameters in an MHD equilibrium which can be approximated by concentric circular flux surfaces. Transport of up to six species of ionized particles, of electron and ion energy, and of poloidal magnetic flux is computed. A wide variety of source terms are calculated including those due to neutral gas, fusion, and auxiliary heating. The code is primarily designed for modeling tokamak plasmas but could be adapted to other toroidal confinement systems

Filterscope diagnostic system on EAST tokamak

International Nuclear Information System (INIS)

Xu, Z.; Wu, Z.W.; Gao, W.; Zhang, L.; Huang, J.; Chen, Y.J.; Wu, C.R.; Zhang, P.F.

2015-01-01

Filterscope diagnostic system, which is designed for monitoring the line emission in fusion plasma has been widely used on fusion devices such as DIII-D, NSTX, CDX-U, KSTAR etc. On EAST (Experimental Advanced Superconducting Tokamak), a filterscope diagnostic system has been mounted to observe the line emission and visible bremsstrahlung emission in plasma from discharge campaign of 2014. It plays a crucial role in studying Edge Localized Modes (ELM) and H-mode, thanks to its high temporal resolution (0.005ms) and good spatial resolution (∼2cm). Furthermore, multi-channel signals at up to 200kHz sampling rates can be digitized simultaneously. The wavelength covers He II (468.5nm), Li I (670.8nm), Li II (548.3nm), C III (465.0nm), O II (441.5nm), Mo I (386.4nm), W I (400.9nm) and visible bremsstrahlung radiation at 538nm besides Dα (656.1nm) and Dγ (433.9nm) with the corresponding wavelength filters. The new developed filterscope system was operating during the EAST 2014 fall experimental campaign and several types ELMs has been observed. (author)

INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

International Nuclear Information System (INIS)

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

2004-03-01

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

The contribution to the energy balance and transport in an advanced-fuel tokamak reactor

International Nuclear Information System (INIS)

Atzeni, S.; Vlad, G.

1985-01-01

The influence of synchrotron radiation emission on the energy balance of an advanced-fuel (such as D- 3 He, or catalyzed-D) tokamak plasma is considered. It is shown that a region in the β-T space exists, where the fusion energy delivered to the plasma overcomes synchrotron and bremsstrahlung energy losses, and which could then allow for ignited operation. 1-Dimensional codes results are also presented, which illustrate the main features of radial transport in a ignited, D- 3 He tokamak plasma

Activation analysis of the compact ignition tokamak

International Nuclear Information System (INIS)

Selcow, E.C.

1986-01-01

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

A study on tokamak fusion reactor - Numerical analyses of MHD equilibrium= and edge plasma transport in tokamak fusion reactor with divertor configurations

Energy Technology Data Exchange (ETDEWEB)

Hong, Sang Hee; Lim, Ki Hang; Kang, Kyung Doo; Ryu, Ji Myung; Kim, Duk Kyu [Seoul National University, Seoul (Korea, Republic of); Cho, Soo Won [Kyungki Unviersity, Suwon (Korea, Republic of)

1995-08-01

In the present project for developing the numerical codes of 2-DMHD equilibrium, edge plasma transport and neutral particle transport for the tokamak plasmas, we compute the plasma equilibrium of double null type and calculate the external coil currents and the plasma parameters used for operation and control data. Also the numerical algorithm is developed to analyse the behavior of edge plasmas in poloidal and radial directions and the programming and debugging of a 2-D transport code are completed. Furthermore, a neutral particle transport code for the edge region is developed and then used for the analysis of the neutral transport phenomena giving the sources in the fluid equations, and expected to supply the input parameters for the edge plasma transport code. 34 refs., 5 tabs., 28 figs. (author)

A distributed high speed data acquisition system for KT5C tokamak

International Nuclear Information System (INIS)

Sun Xiang; Wang Zhijiang; Lu Ronghua; Wang Jun; Yu Yi; Zhu Zhenghua; Wen Yizhi; Wan Shude; Liu Wandong; Yu Changxuan

2005-01-01

The development of a distributed data acquisition system with low cost to implement high speed data collection through the campus networks for a small tokamak, KT5C, is presented. Data of 512 k bytes at 5 MHz from 5 channels for each can be collected during about 10s after three researchers at different positions demand this system for acquisitions. This system realizes long distance multiuser operations; virtually efficiency of the data acquisition is enhanced. (authors)

From profile to sawtooth control: developing feedback control using ECRH/ECCD systems on the TCV tokamak

International Nuclear Information System (INIS)

Paley, J I; Felici, F; Coda, S; Goodman, T P

2009-01-01

Real time control of heating systems is essential to maximize plasma performance and avoid or neutralize instabilities under changing plasma conditions. Several feedback control algorithms have been developed on the Tokamak a Configuration Variable (TCV) tokamak that use the electron cyclotron (ECRH/ECCD) system to control a wide range of plasma properties, including the plasma current, shape, profiles as well as the sawtooth instability. Controllers have been developed to obtain sawteeth of a pre-determined period, to maximize the sawtooth period using an extremum seeking control algorithm and finally to provide simultaneous control of the plasma emission profile peak and width using multiple independent EC actuators.

Magneto-hydro-dynamic simulation of Edge-Localised-Modes in tokamaks

International Nuclear Information System (INIS)

Pamela, S.

2010-01-01

In order to produce energy from fusion reactions in a tokamak, the plasma must reach temperatures higher than that of our sun. The operation regime called H-mode enables one to acquire a plasma confinement close to fusion conditions. Due to the formation of a transport barrier at the plasma edge, turbulent transport is reduced, and the total plasma pressure increases, resulting in a strong pressure gradient at the edge. If this pressure gradient, localised at the plasma-vacuum boundary, becomes too steep, a magneto-hydro-dynamic instability is triggered and part of the plasma pressure is lost. This instability, hence called Edge-Localised-Mode, provokes large heat fluxes on some of the plasma-facing components of the machine, which could set operational limits for a tokamak the size of ITER. In order to understand this instability, and to determine the non-linear mechanisms behind the ELMs, the JOREK code is used. The work presented in this thesis is based on MHD simulations of ballooning modes (responsible for ELMs) with the JOREK code. At first, simulations are done for standard plasmas, inspired of experimental machines. In particular, the plasma rotation at equilibrium, in the region of the edge pressure gradient, is studied in order to obtain an analysis of the effects that such a rotation can have on the linear stability of ELMs and on their non-linear evolution. Then, as a second step, simulations are applied to plasmas of the experimental tokamaks JET and MAST (England). This permits the direct comparison of simulation results with experimental observations, with the main goal of improving our global understanding of ELMs. Adding to this physics aspect, the confrontation of the JOREK code with diagnostics of JET and MAST brings to a validation of simulations, which should prove that the simulations which were obtained do correspond to ELM instabilities. This first step towards the validation of the code is crucial concerning the simulation of ELMs in ITER

Theoretical issues in tokamak confinement: (i) Internal/edge transport barriers and (ii) runaway avalanche confinement

International Nuclear Information System (INIS)

Connor, J.W.; Helander, P.; Thyagaraja, A.; Andersson, F.; Fueloep, T.; Eriksson, L.-G.; Romanelli, M.

2001-01-01

This paper summarises a number of distinct, but related, pieces of work on key confinement issues for tokamaks, in particular the formation of internal and edge transport barriers, both within turbulent and neoclassical models, and radial diffusion of avalanching runaway electrons. First-principle simulations of tokamak turbulence and transport using the two-fluid, electromagnetic, global code CUTIE are described. The code has demonstrated the spontaneous formation of internal transport barriers near mode rational surfaces, in qualitative agreement with observations on JET and RTP. The theory of neoclassical transport in an impure, toroidal plasma has been extended to allow for steeper pressure and temperature gradients than are usually considered, and is then found to become nonlinear under conditions typical of the tokamak edge. For instance, the particle flux is found to be a nonmonotonic function of the gradients, thus allowing for a bifurcation in the ion particle flux. Finally, it is shown that radial diffusion caused by magnetic fluctuations can effectively suppress avalanches of runaway electrons if the fluctuation amplitude exceeds δB/B∼10 -3 . (author)

Performance Projections For The Lithium Tokamak Experiment (LTX)

International Nuclear Information System (INIS)

Majeski, R.L.; Berzak, T.; Gray, R.; Kaita, T.; Kozub, F.; Levinton, D.P.; Lundberg, J.; Manickam, G.V.; Pereverzev, K.; Snieckus, V.; Soukhanovskii, J.; Spaleta, D.; Stotler, T.; Strickler, J.; Timberlake, J.; Zakharov, L.; Zakharov, Y.

2009-01-01

Use of a large-area liquid lithium limiter in the CDX-U tokamak produced the largest relative increase (an enhancement factor of 5-10) in Ohmic tokamak confinement ever observed. The confinement results from CDX-U do not agree with existing scaling laws, and cannot easily be projected to the new lithium tokamak experiment (LTX). Numerical simulations of CDX-U low recycling discharges have now been performed with the ASTRA-ESC code with a special reference transport model suitable for a diffusion-based confinement regime, incorporating boundary conditions for nonrecycling walls, with fueling via edge gas puffing. This model has been successful at reproducing the experimental values of the energy confinement (4-6 ms), loop voltage (<0.5 V), and density for a typical CDX-U lithium discharge. The same transport model has also been used to project the performance of the LTX, in Ohmic operation, or with modest neutral beam injection (NBI). NBI in LTX, with a low recycling wall of liquid lithium, is predicted to result in core electron and ion temperatures of 1-2 keV, and energy confinement times in excess of 50 ms. Finally, the unique design features of LTX are summarized

Joint research using small tokamaks

International Nuclear Information System (INIS)

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

2005-01-01

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

Axisymmetric stability of vertically asymmetric Tokamaks at large beta poloidal

Energy Technology Data Exchange (ETDEWEB)

Yamazaki, K.; Fishman, H.; Okabayashi, M. (Princeton Univ., NJ (USA). Plasma Physics Lab.); Todd, A.M.M. (Grumman Aerospace Corp., Princeton, NJ (USA))

1983-11-01

The rigid-mode stability of high-..beta.. vertically asymmetric Tokamak equilibria with quasi-uniform current profile is investigated analytically using toroidicity-shaping double expansion method. It is found that vertical stability at large beta poloidal is mainly determined by a coupling between the shape of the plasma surface and the Shafranov shift of the magnetic axis. To the lowest order, symmetric components of the plasma surface shape are found to be the critical destabilizing elements. Asymmetric components have little effect. The inclusion of higher order terms in the high-..beta.. Tokamak expansion leads to further destabilization. These analytic insights are qualitatively confirmed by numerical stability calculations using the PEST code with parabolic safety-factor profile.

A Distributed Synchronization and Timing System on the EAST Tokamak

Science.gov (United States)

Luo, Jiarong; Wu, Yichun; Shu, Yantai

2008-08-01

A key requirement for the EAST distributed control system (EASTDCS) is time synchronization to an accuracy of RTOS). The DSTS provides the control and the data acquisition systems with reference clocks (0.01 Hz 10 MHz) and delayed trigger times ( 1 mus 4294 s). These are produced by a Core Module Unit (CMU) connected by optical fibres to many Local Synchronized Node Units (LSNU). The fibres provide immunity from electrical noise and are of equal length to match clock and trigger delays between systems. This paper describes the architecture of the DSTS on the EAST tokamak and provides an overview of the characteristics of the main and local units.

Fast-wave current drive modelling for large non-circular tokamaks

International Nuclear Information System (INIS)

Batchelor, D.B.; Goldfinger, R.C.; Jaeger, E.F.; Carter, M.D.; Swain, D.W.; Ehst, D.; Karney, C.F.F.

1990-01-01

It is widely recognized that a key element in the development of an attractive tokamak reactor, and in the successful achievement of the mission of ITER, is the development of an efficient steady-state current drive technique. Fast waves in the ion cyclotron range of frequencies hold the promise to drive steady-state currents with the required efficiency and to effectively heat the plasma to ignition. Advantages over other heating and current drive techniques include low cost per watt and the ability to penetrate to the center of high-density plasmas. The primary issues that must be resolved are: can an antenna array be designed to radiate the required spectrum of waves and have adequate coupling properties? Will the rf power be efficiently absorbed by electrons in the desired velocity range without unacceptable parasitic damping by fuel ions or α particles? What will the efficiency of current drive be when toroidal effects such as trapped particles are included? Can a practical rf system be designed and integrated into the device? We have addressed these issues by performing extensive calculations with ORION, a 2-D code, and the ray tracing code RAYS, which calculate wave propagation, absorption and current drive in tokamak geometry, and with RIP, a 2-D code that self-consistently calculates current drive with MHD equilibrium. An important figure of merit in this context is the integrated, normalized current drive efficiency. The calculations that we present here emphasize the ITER device. We consider a low-frequency scenario such that no ion resonances appear in the machine, and a high-frequency scenario such that the deuterium second harmonic resonance is just outside the plasma and the tritium second harmonic is in the plasma, midway between the magnetic axis and the inside edge. In both cases electron currents are driven by combined TTMP and Landau damping of the fast waves

Conceptual design of a commercial tokamak reactor using resistive magnets

International Nuclear Information System (INIS)

LeClaire, R.J. Jr.

1988-01-01

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

MTX [Microwave Tokamak Experiment] diagnostic and auxiliary systems for confinement, transport, and plasma physics studies

International Nuclear Information System (INIS)

Hooper, E.B.; Allen, S.L.; Casper, T.A.; Thomassen, K.I.

1989-01-01

This note describes the diagnostics and auxiliary systems on the Microwave Tokamak Experiment (MTX) for confinement, transport, and other plasma physics studies. It is intended as a reference on the installed and planned hardware on the machine for those who need more familiarity with this equipment. Combined with the tokamak itself, these systems define the opportunities and capabilities for experiments in the MTX facility. We also illustrate how these instruments and equipment are to be used in carrying out the MTX Operations Plan. Near term goals for MTX are focussed on the absorption and heating by the microwave beam from the FEL, but the Plan also includes using the facility to study fundamental phenomena in the plasma, to control MHD activity, and to drive current noninductively

Testing and commissioning the multinode ECRH realtime control system on the FTU tokamak

International Nuclear Information System (INIS)

Galperti, C.; Boncagni, L.; Alessi, E.; Sozzi, C.; Nowak, S.; Granucci, G.; Minelli, D.; Marchetto, C.

2014-01-01

Highlights: • We conceived, developed and commissioned a distributed multinode control hardware with proven real-time performances. • The adopted hardware solution is modular and reconfigurable. • The adopted software solution is able to host many experimental scenarios and is totally remotely programmable, configurable and testable. • Effective results in on-line MHD instability detection are presented. - Abstract: In tokamak machines, the ECRH heating system is crucial for plasma heating and for stability control. To be reliable, an ECRH control system should be deeply integrated into the supervision and control systems of the machine, and must be interconnected to the diagnostic instruments and the power actuators of the plant. Moreover, several ECRH experiments are under investigation by the community. So, for the sake of efficiency, it should be possible to reprogram a control system on the fly and possibly from remote locations, even during experiment campaigns. This paper presents the new ECRH control system under development at the FTU tokamak. This system consists of multiple units that acquire and process data and are linked through Ethernet and dedicated fiber-optic data links, under a Linux/MARTe framework. This paper also presents open-loop operative results, both about performances of the control system and about signal processing of the diagnostics relevant to MHD control

Testing and commissioning the multinode ECRH realtime control system on the FTU tokamak

Energy Technology Data Exchange (ETDEWEB)

Galperti, C., E-mail: galperti@ifp.cnr.it [EURATOM – ENEA – CNR Fusion Association, CNR-IFP, via R. Cozzi 53, 20125 Milan (Italy); Boncagni, L., E-mail: luca.boncagni@enea.it [EURATOM – ENEA – CNR Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Alessi, E.; Sozzi, C.; Nowak, S.; Granucci, G.; Minelli, D.; Marchetto, C. [EURATOM – ENEA – CNR Fusion Association, CNR-IFP, via R. Cozzi 53, 20125 Milan (Italy)

2014-03-15

Highlights: • We conceived, developed and commissioned a distributed multinode control hardware with proven real-time performances. • The adopted hardware solution is modular and reconfigurable. • The adopted software solution is able to host many experimental scenarios and is totally remotely programmable, configurable and testable. • Effective results in on-line MHD instability detection are presented. - Abstract: In tokamak machines, the ECRH heating system is crucial for plasma heating and for stability control. To be reliable, an ECRH control system should be deeply integrated into the supervision and control systems of the machine, and must be interconnected to the diagnostic instruments and the power actuators of the plant. Moreover, several ECRH experiments are under investigation by the community. So, for the sake of efficiency, it should be possible to reprogram a control system on the fly and possibly from remote locations, even during experiment campaigns. This paper presents the new ECRH control system under development at the FTU tokamak. This system consists of multiple units that acquire and process data and are linked through Ethernet and dedicated fiber-optic data links, under a Linux/MARTe framework. This paper also presents open-loop operative results, both about performances of the control system and about signal processing of the diagnostics relevant to MHD control.

Tokamak poloidal-field systems. Progress report, January 1-December 31, 1981

International Nuclear Information System (INIS)

Rogers, J.D.

1982-03-01

Work on the superconducting tokamak poloidal field system (TPFS) program is being redirected. The development of the 20 MJ, 50 kA, 7.5 T superconducting programmed energy storage coil is being terminated. The superconductor for the 20 MJ coil is being processed only to an intermediate state, and manufacture of the epoxy fiberglass dewar is being stopped. Further, development of the TPFS test facility is in abeyance. Change in program emphasis arises from prospective rf plasma current driven or beam heated tokamaks with programmed coil characteristics for the poloidal field being different from those to have been simulated by the 20 MJ coil and from budgetary constraints. Work is reported on the development of the coil, conductor, nonconducting dewar, and test facility to the recent time when the program change was instigated. Work in support of the Large Coil Test Facility (LCTF) and the Fusion Engineering Design (FED) Center is given. Analysis of the experiments on the 400 kJ METS coil test was completed

Joint research using small tokamaks

International Nuclear Information System (INIS)

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

2005-01-01

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

Millimeter-wave imaging diagnostics systems on the EAST tokamak (invited)

Energy Technology Data Exchange (ETDEWEB)

Zhu, Y. L.; Xie, J. L., E-mail: jlxie@ustc.edu.cn; Yu, C. X.; Zhao, Z. L.; Gao, B. X.; Chen, D. X.; Liu, W. D.; Liao, W.; Qu, C. M.; Luo, C. [School of Physics, University of Science and Technology of China, Anhui 230026 (China); Hu, X.; Spear, A. G.; Luhmann, N. C.; Domier, C. W.; Chen, M.; Ren, X. [University of California, Davis, California 95616 (United States); Tobias, B. J. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

2016-11-15

Millimeter-wave imaging diagnostics, with large poloidal span and wide radial range, have been developed on the EAST tokamak for visualization of 2D electron temperature and density fluctuations. A 384 channel (24 poloidal × 16 radial) Electron Cyclotron Emission Imaging (ECEI) system in F-band (90-140 GHz) was installed on the EAST tokamak in 2012 to provide 2D electron temperature fluctuation images with high spatial and temporal resolution. A co-located Microwave Imaging Reflectometry (MIR) will be installed for imaging of density fluctuations by December 2016. This “4th generation” MIR system has eight independent frequency illumination beams in W-band (75-110 GHz) driven by fast tuning synthesizers and active multipliers. Both of these advanced millimeter-wave imaging diagnostic systems have applied the latest techniques. A novel design philosophy “general optics structure” has been employed for the design of the ECEI and MIR receiver optics with large aperture. The extended radial and poloidal coverage of ECEI on EAST is made possible by innovations in the design of front-end optics. The front-end optical structures of the two imaging diagnostics, ECEI and MIR, have been integrated into a compact system, including the ECEI receiver and MIR transmitter and receiver. Two imaging systems share the same mid-plane port for simultaneous, co-located 2D fluctuation measurements of electron density and temperature. An intelligent remote-control is utilized in the MIR electronics systems to maintain focusing at the desired radial region even with density variations by remotely tuning the probe frequencies in about 200 μs. A similar intelligent technique has also been applied on the ECEI IF system, with remote configuration of the attenuations for each channel.

Tokamaks. 2. ed.

International Nuclear Information System (INIS)

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

1997-01-01

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

Tokamak devices: towards controlled fusion

International Nuclear Information System (INIS)

Trocheris, M.

1975-01-01

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

Numerical studies of transport processes in Tokamak plasma

International Nuclear Information System (INIS)

Spineanu, F.; Vlad, M.

1984-09-01

The paper contains the summary of a set of studies of the transport processes in tokamak plasma, performed with a one-dimensional computer code. The various transport models (which are implemented by the expressions of the transport coefficients) are presented in connection with the regimes of the dynamical development of the discharge. Results of studies concerning the skin effect and the large scale MHD instabilities are also included

Control system of neoclassical tearing modes in real time on HL-2A tokamak

Science.gov (United States)

Yan, Longwen; Ji, Xiaoquan; Song, Shaodong; Xia, Fan; Xu, Yuan; Ye, Jiruo; Jiang, Min; Chen, Wenjin; Sun, Tengfei; Liang, Shaoyong; Ling, Fei; Ma, Rui; Huang, Mei; Qu, Hongpeng; Song, Xianming; Yu, Deliang; Shi, Zhongbin; Liu, Yi; Yang, Qingwei; Xu, Min; Duan, Xuru; Liu, Yong

2017-11-01

The stability and performance of tokamak plasmas are routinely limited by various magneto-hydrodynamic instabilities, such as neoclassical tearing modes (NTMs). This paper presents a rather simple method to control the NTMs in real time (RT) on a tokamak, including the control principle of a feedback approach for RT suppression and stabilization for the NTMs. The control system combines Mirnov, electron cyclotron emission, and soft X-ray diagnostics used for determining the NTM positions. A methodology for fast detection of 2/1 or 3/2 NTM positions with 129 × 129 grid reconstruction is elucidated. The forty poloidal angles for steering the electron cyclotron resonance heating (ECRH)/electron cyclotron current drive launcher are used to establish the alignment of antenna mirrors with the center of the NTM and to ensure launcher emission intersecting with the rational surface of a magnetic island. Pilot experiments demonstrate the RT control capability to trace the conventional tearing modes (CTMs) in the HL-2A tokamak. The 2/1 CTMs have been suppressed or stabilized by the ECRH power deposition on site or with the steerable launcher.

Development of remote control integrator system on Tokamak

International Nuclear Information System (INIS)

Wu Yichun; Wang Lingzhi; Shu Shuangbao

2014-01-01

In order to meet with the requirement of electromagnetic diagnosis to the J-TEXT Tokamak, a remote control integrator system was developed. With modular design method, the integrator system is composed of the integrator cards, a control card, a linear power card and the BNC interface cards, and it uses the PC control soft- ware to conduct network control. An integrator system provides 32 integrator channels, and all integral channels have four kinds of integral time constants for remote selection and provide three kinds of integrator running control methods. According to laboratory and J-TEXT field testing, it shows that the output voltage range is -10-10 V, output noise is not more than 5 mV, and for the four kinds of integral time constants, the integral output drifts are all less than 5 mV within 100 s for each integrator channel. (authors)

The ARIES tokamak fusion reactor study

International Nuclear Information System (INIS)

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

1989-01-01

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

First experiments with SST-1 tokamak

International Nuclear Information System (INIS)

Saxena, Y.C.

2005-01-01

SST-1, a steady state superconducting tokamak, is undergoing commissioning tests at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in a tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. An Ohmic transformer is provided for plasma breakdown and initial current ramp up. SST-1 deploys a fully welded ultra high vacuum vessel. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. The auxiliary current drive is based on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Frequency system (ICRF) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating at 84 GHz and a Neutral Beam Injection (NBI) system with peak power of 0.8 MW (at 80 keV) with variable beam energy in range of 10-80 keV. The ICRF system would also be used for initial breakdown and wall conditioning experiments. Detailed commissioning tests on the cryogenic system and experiments on the hydraulic characters and cool down features of single TF coils have been completed prior to the cool down of the entire superconducting system. Results of the single TF magnet cool down, and testing of the magnet system are presented. First experiments related to the breakdown and the current ramp up will subsequently be carried out. (author)

GATO: an MHD stability code for axisymmetric plasmas with internal separatrices

International Nuclear Information System (INIS)

Bernard, L.C.; Helton, F.J.; Moore, R.W.

1981-07-01

The GATO code computes the growth rate of ideal magnetohydrodynamic instabilities in axisymmetric geometries with internal separatrices such as doublet and expanded spheromak. The basic method, which uses a variational principle and a Galerkin procedure to obtain a matrix eigenvalue problem, is common to the ERATO and PEST codes. A new coordinate system has been developed to handle the internal separatrix. Efficient algorithms have been developed to solve the matrix eigenvalue problem for matrices of rank as large as 40,000. Further improvement is expected using graph theoretical techniques to reorder the matrices. Using judicious mesh repartition, the marginal point can be determined with great precision. The code has been extensively used to optimize doublet and general tokamak plasmas

SASSYS LMFBR systems code

International Nuclear Information System (INIS)

Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

1983-01-01

The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

Recent advances in the HL-2A tokamak experiments

International Nuclear Information System (INIS)

Liu, Y.; Ding, X.T.; Yang, Q.W.; Yan, L.W.; Liu, D.Q.; Xuan, W.M.; Chen, L.Y.; Song, X.M.; Cao, Z.; Zhang, J.H.; Mao, W.C.; Zhou, C.P.; Li, X.D.; Wang, S.J.; Yan, J.C.; Bu, M.N.; Chen, Y.H.; Cui, C.H.; Cui, Z.Y.; Deng, Z.C.; Hong, W.Y.; Hu, H.T.; Huang, Y.; Kang, Z.H.; Li, B.; Li, W.; Li, F.Z.; Li, G.S.; Li, H.J.; Li, Q.; Li, Y.G.; Li, Z.J.; Liu, Yi; Liu, Z.T.; Luo, C.W.; Mao, X.H.; Pan, Y.D.; Rao, J.; Shao, K.; Song, X.Y.; Wang, M.; Wang, M.X.; Wang, Q.M.; Xiao, Z.G.; Xie, Y.F.; Yao, L.H.; Yao, L.Y.; Zheng, Y.J.; Zhong, G.W.; Zhou, Y.; Pan, C.H.

2005-01-01

Two experiment campaigns were conducted on the HL-2A tokamak in 2003 and 2004 after the first plasma was obtained at the end of 2002. Progresses in many aspects have been made, especially in the divertor discharge and feedback control of plasma configuration. Up to now, the following operation parameters have been achieved: I p = 320 kA, B t = 2.2 T and discharge duration T d = 1580 ms. With the feedback control of plasma current and horizontal position, an excellent repeatability of the discharge has been achieved. The tokamak has been operated at both limiter configuration and single null (SN) divertor configuration. The HL-2A SN divertor configuration is simulated with the MHD equilibrium code SWEQU. When the divertor configuration is formed, the impurity radiation in the main plasma decreases remarkably

Finite pressure effects on the tokamak sawtooth crash

International Nuclear Information System (INIS)

Nishimura, Yasutaro

1998-07-01

The sawtooth crash is a hazardous, disruptive phenomenon that is observed in tokamaks whenever the safety factor at the magnetic axis is below unity. Recently, Tokamak Test Fusion Reactor (TFTR) experimental data has revealed interesting features of the dynamical pressure evolution during the crash phase. Motivated by the experimental results, this dissertation focuses on theoretical modeling of the finite pressure effects on the nonlinear stage of the sawtooth crash. The crash phase has been studied numerically employed a toroidal magnetohydrodynamic (MHD) initial value code deduced from the FAR code. For the first time, by starting from a concentric equilibrium, it has been shown that the evolution through an m/n = 1/1 magnetic island induces secondary high-n ballooning instabilities. The magnetic island evolution gives rise to convection of the pressure inside the inversion radius and builds up a steep pressure gradient across the island separatrix, or current sheet, and thereby triggers ballooning instabilities below the threshold for the axisymmetric equilibrium. Due to the onset of secondary ballooning modes, concomitant fine scale vortices and magnetic stochasticity are generated. These effects produce strong flows across the current sheet, and thereby significant modify the m = 1 driven magnetic reconnection process. The resultant interaction of the high-n ballooning modes with the magnetic reconnection process is discussed

Economic considerations of commercial tokamak options

International Nuclear Information System (INIS)

Dabiri, A.E.

1986-05-01

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

Prospects for Tokamak Fusion Reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.

1995-01-01

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

Vectorization, parallelization and porting of nuclear codes (porting). Progress report fiscal 1998

International Nuclear Information System (INIS)

Nemoto, Toshiyuki; Kawai, Wataru; Ishizuki, Shigeru; Kawasaki, Nobuo; Kume, Etsuo; Adachi, Masaaki; Ogasawara, Shinobu

2000-03-01

Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system, the AP3000 system and the Paragon system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 12 codes in fiscal 1998. These results are reported in 3 parts, i.e., the vectorization and parallelization on vector processors part, the parallelization on scalar processors part and the porting part. In this report, we describe the porting. In this porting part, the porting of Monte Carlo N-Particle Transport code MCNP4B2 and Reactor Safety Analysis code RELAP5 on the AP3000 are described. In the vectorization and parallelization on vector processors part, the vectorization of General Tokamak Circuit Simulation Program code GTCSP, the vectorization and parallelization of Molecular Dynamics Ntv Simulation code MSP2, Eddy Current Analysis code EDDYCAL, Thermal Analysis Code for Test of Passive Cooling System by HENDEL T2 code THANPACST2 and MHD Equilibrium code SELENEJ on the VPP500 are described. In the parallelization on scalar processors part, the parallelization of Monte Carlo N-Particle Transport code MCNP4B2, Plasma Hydrodynamics code using Cubic Interpolated propagation Method PHCIP and Vectorized Monte Carlo code (continuous energy model/multi-group model) MVP/GMVP on the Paragon are described. (author)

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

Design of FPGA based high-speed data acquisition and real-time data processing system on J-TEXT tokamak

International Nuclear Information System (INIS)

Zheng, W.; Liu, R.; Zhang, M.; Zhuang, G.; Yuan, T.

2014-01-01

Highlights: • It is a data acquisition system for polarimeter–interferometer diagnostic on J-TEXT tokamak based on FPGA and PXIe devices. • The system provides a powerful data acquisition and real-time data processing performance. • Users can implement different data processing applications on the FPGA in a short time. • This system supports EPICS and has been integrated into the J-TEXT CODAC system. - Abstract: Tokamak experiment requires high-speed data acquisition and processing systems. In traditional data acquisition system, the sampling rate, channel numbers and processing speed are limited by bus throughput and CPU speed. This paper presents a data acquisition and processing system based on FPGA. The data can be processed in real-time before it is passed to the CPU. It provides processing ability for more channels with higher sampling rates than the traditional data acquisition system while ensuring deterministic real-time performance. A working prototype is developed for the newly built polarimeter–interferometer diagnostic system on the Joint Texas Experimental Tokamak (J-TEXT). It provides 16 channels with 120 MHz maximum sampling rate and 16 bit resolution. The onboard FPGA is able to calculate the plasma electron density and Faraday rotation angel. A RAID 5 storage device is adopted providing 700 MB/s read–write speed to buffer the data to the hard disk continuously for better performance

Tokamak plasma equilibrium problems with anisotropic pressure and rotation and their numerical solution

International Nuclear Information System (INIS)

Ivanov, A. A.; Martynov, A. A.; Medvedev, S. Yu.; Poshekhonov, Yu. Yu.

2015-01-01

In the MHD tokamak plasma theory, the plasma pressure is usually assumed to be isotropic. However, plasma heating by neutral beam injection and RF heating can lead to a strong anisotropy of plasma parameters and rotation of the plasma. The development of MHD equilibrium theory taking into account the plasma inertia and anisotropic pressure began a long time ago, but until now it has not been consistently applied in computational codes for engineering calculations of the plasma equilibrium and evolution in tokamak. This paper contains a detailed derivation of the axisymmetric plasma equilibrium equation in the most general form (with arbitrary rotation and anisotropic pressure) and description of the specialized version of the SPIDER code. The original method of calculation of the equilibrium with an anisotropic pressure and a prescribed rotational transform profile is proposed. Examples of calculations and discussion of the results are also presented

Integrated Tokamak modeling: When physics informs engineering and research planning

Science.gov (United States)

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. It discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.

Neoclassical alpha-particle losses in tokamaks allowing for large orbit widths

International Nuclear Information System (INIS)

Cox, M.; O'Brien, M.R.; Zaitsev, F.S.

1994-01-01

Alpha-particle physics is of particular importance now that research into controlled fusion has reached thermonuclear parameters and D-T fuel has been used in JET and TFTR. Here we address the important topic of α-particle transport: if transport is too low helium ash accumulates quenching the burn; if it is too high heating of the plasma by fast α-particles is insufficient to maintain the burn. We give results from simulations of α-particle distributions (f α ) which self-consistently treat α-particle birth, collisional slowing down and neoclassical radial transport. The (steady-state) f α is calculated by the FPP code as a function of speed (v), pitch-angle (θ) and flux surface radius (r). This code is based on a 3D Fokker-Planck theory of 'banana regime' neoclassical effects in tokamaks which can treat large deviations of fast ion orbits from flux surfaces and non-Maxwellian distributions. The code reproduces standard neoclassical results for Maxwellian distributions in the large aspect ratio (ε) and small orbit width (Δ) limits (e.g. radial fluxes, conductivities and bootstrap currents), but can also be used for small ε and large Δ which are difficult to treat analytically. The code is particularly useful for α-particle studies as (a) the experimental evidence is that fast ion transport is usually consistent with neoclassical theory, unlike electron or thermal ion transport, and (b) trapped fast ion orbits can deviate greatly from flux surfaces. An alternative to this Fokker-Planck treatment is Monte Carlo modelling. However, representation of the detailed structure of f α (θ,v,r) would require very large number of particles, and hence be very slow. Calculations have been made for parameters typical of TFTR, JET, SSTR (an 'advanced tokamak' reactor) and STR (a tight aspect ratio or 'spherical' tokamak reactor, though only the JET results are discussed in detail. (author) 4 refs., 4 figs

Transport simulations of a density limit in radiation-dominated tokamak discharges: Profile effects

International Nuclear Information System (INIS)

Stotler, D.P.

1988-06-01

The density limit observed in tokamak experiments is thought to be due to a radiative collapse of the current channel. A transport code coupled with an MHD equilibrium routine is used to determine the detailed, self-consistent evolution of the plasma profiles in tokamak discharges with radiated power close to or equalling the input power. The present work is confined to ohmic discharges in steady state. It is found that the shape of the density profile can have a significant impact on the variation of the maximum electron density with plasma current. Analytic calculations confirm this result. 41 refs., 9 figs

Transport simulations of a density limit in radiation-dominated tokamak discharges: profile effects

International Nuclear Information System (INIS)

Stotler, D.P.

1988-01-01

The density limit observed in tokamak experiments is thought to be due to a radiative collapse of the current channel. A transport code coupled with a magnetohydrodynamic (MHD) equilibrium routine is used to determine the detailed, self-consistent evolution of the plasma profiles in tokamak discharges with radiated power close to or equaling the input power. The present work is confined to Ohmic discharges in steady state. It is found that the shape of the density profile can have a significant impact on the variation of the maximum electron density with plasma current. Analytic calculations confirm this result

Tokamak Physics Experiment (TPX) design

International Nuclear Information System (INIS)

Schmidt, J.A.

1995-01-01

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

Simulations of Turbulence in Tokamak Edge and Effects of Self-Consistent Zonal Flows

Science.gov (United States)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress is reported on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge. This extends previous work to include self-consistent zonal flows and their effects. The previous work addressed simulation of L-mode tokamak edge turbulence using the turbulence code BOUT that solves Braginskii-based plasma fluid equations in tokamak edge domain. The calculations use realistic single-null geometry and plasma parameters of the DIII-D tokamak and produce fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the U.S. Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Discrete Alfven waves in the TORTUS tokamak

International Nuclear Information System (INIS)

Amagishi, Y.; Ballico, M.J.; Cross, R.C.; Donnely, I.J.

1989-01-01

Discrete Alfven Waves (DAWs) have been observed as antenna resistance peaks and as enhanced edge fields in the TORTUS tokamak during Alfven wave heating experiments. A kinetic theory code has been used to calculate the antenna loading and the structure of the DAW fields for a range of plasma current and density profiles. There is fair agreement between the measured and predicted amplitude of the DAW fields in the plasma edge when both are normalized to the same antenna power

Assessment of feasibility of helium ash exhaust and heat removal by pumped-limiter in tokamak fusion reactor

International Nuclear Information System (INIS)

Hitoki, Shigehisa; Sugihara, Masayoshi; Saito, Seiji; Fujisawa, Noboru

1985-01-01

A detailed calculation of the behavior of fuel and He particles in tokamak reactor with pumped-limiter is performed by one-dimensional tokamak transport code. Energy of neutral particles flowing back from limiter chamber is calculated by two-dimensional Monte Carlo neutral code. Feasibility of He ash exhaust and heat removal by the pumped-limiter are analyzed. Following features of the pumped-limiter are clarified: (1) Electron temperature decays rapidly in radial direction in scrape-off layer, while density profile is broader than that of temperature. (2) Helium accumulation in main plasma can be kept at desired level by rather short limiter and moderate pumping system. (3) Minimum amount of tritium pumped out little depends on limiter length. (4) Although high temperature plasma in scrape-off layer could be realized by large pumping and ideal pellet injection, it is not sufficiently high to reduce the erosion of the limiter surface and the leading edge. In conclusion, He ash exhaust may be possible by the pumped-limiter, while the heat load and erosion will be so high that the pumped-limiter may not be applicable unless the boundary plasma is cooled by radiation or by some other means. (author)

Stability of tokamak magnetic configuration with a poloidal divertor

International Nuclear Information System (INIS)

Bazaeva, A.V.; Bykov, V.E.; Georgievskii, A.V.; Kaminskii, A.O.; Peletminskaya, V.G.; Pyatov, V.H.

1979-02-01

This paper investigates instabilities in the preseparatrix region of a tokamak magnetic configuration with a poloidal divertor with respect to perturbations produced by various irregularities in the manufacturing of tokamak magnetic systems. A computer solution, a system of differential equations describing the behavior of a force line, showed that small perturbation amplitudes may be the cause of the stochastic instability of force lines in the preseparatrix region. This instability is responsible for a number of demands on the accuracy in the manufacturing of tokamak magnetic systems. In particular, the misalignment in the divertor ring must not be larger than 0.5 0 , its displacement must be less than Δ/R = 10 -2 (Δ/R -2 ). This study can be used in the design of large thermonuclear installations

Acceleration mechanism of vertical displacement event and its amelioration in tokamak disruptions

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Yoshino, Ryuji; Pomphrey, N.; Jardin, S.C.

1996-01-01

Vertical displacement events (VDEs), which are frequently observed in disruptive discharges of elongated tokamaks, are investigated using the Tokamak Simulation Code. We show that disruption events such as a sudden plasma pressure drop (β p collapse) and the subsequent plasma current quench (I p quench) can accelerate VDEs due to the adverse destabilizing effect of the resistive shell, which has previously been thought to stabilize VDEs. In a tokamak with a surrounding shell which is asymmetric with respect to the geometric midplane, the I p quench also causes an additional VDE acceleration due to the vertical imbalance of the attractive force. While the shell-geometry characterizes the VDE dynamics, the growth rate of VDEs depends strongly on the magnitude of the β p collapse, the speed of the I p quench and the n-index of the plasma equilibrium just before the disruption. An amelioration of I p quench-induced VDEs was experimentally established in the JT-60U tokamak by optimizing the vertical location of the plasma just prior to the disruption. The JT-60U vacuum vessel is shown to be suitable for preventing the β p collapse-induced VDE. (author)

TPX tokamak construction management

International Nuclear Information System (INIS)

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

1995-01-01

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

DEALS: a maintainable superconducting magnet system for tokamak fusion reactors

International Nuclear Information System (INIS)

Hseih, S.Y.; Danby, G.; Powell, J.R.

1979-01-01

The feasibility of demountable superconducting magnet systems has been examined in a design study of a DEALS [Demountable Externally Anchored Low Stress] TF magnet for an HFITR [High Field Ignition Test Reactor] Tokamak device. All parts of the system appear feasible, including the demountable superconducting joints. Measurements on small scale prototype joints indicate that movable pressure contact joints exhibit acceptable electrical, mechanical, and cryogenic performance. Such joints permit a relatively simple support structure and are readily demountable. Assembly and disassembly sequences are described whereby any failed portion of the magnet, or any part of the reactor inside the TF coils can be removed and replaced if necessary

Performance and analysis of the TVTS diagnostic system on HT-7 tokamak

International Nuclear Information System (INIS)

Han Xiaofeng; Shao Chunqiang; Xi Xiaoqi; Zhao Junyu; Qing Zang; Yang Jianhua; Dai Xingxing

2013-01-01

A high spatial resolution imaging Thomson scattering diagnostic system was developed in ASIPP. After about one month trial running on the superconducting HT-7 tokamak, the system was proved to be capable of measuring plasma electron temperature. The system setup and data calibration are described in this paper and then the instrument function is studied in detail, as well as the measurement capability, an electron temperature of 50 eV to 2 keV and density beyond 1x10"1"9 m"-"3. Finally, the data processing method and experimental results are presented. (author)

Physics design of an ultra-long pulsed tokamak reactor

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

Saxena, Y.C.

2000-01-01

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

Effects of classical and neo-classical cross-field transport of tungsten impurity in realistic tokamak geometry

Energy Technology Data Exchange (ETDEWEB)

Yamoto, S.; Inoue, H.; Sawada, Y.; Hatayama, A. [Faculty of Science and Technology, Keio University, Yokohama (Japan); Homma, Y.; Hoshino, K. [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Bonnin, X. [ITER Organization, St. Paul Lez Durance (France); Coster, D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Schneider, R. [Ernst-Moritz-Arndt University Greifswald (Germany)

2016-08-15

The initial simulation study of the neoclassical perpendicular self-diffusion transport in the SOL/Divertor regions for a realistic tokamak geometry with the IMPGYRO code has been performed in this paper. One of the most unique features of the IMPGYRO code is calculating exact Larmor orbit of the test particle instead of assuming guiding center approximation. Therefore, effects of the magnetic drifts in realistic tokamaks are naturally taken into account in the IMPGYRO code. This feature makes it possible to calculate neoclassical transport processes, which possibly become large in the SOL/divertor plasma. Indeed, neoclassical self-diffusion process, the resultant effect of the combination of magnetic drift and Coulomb collisions with background ions, has already been included in the IMPGYRO model. In the present paper, prior to implementing the detailed model of neoclassical transport process into IMPGYRO, we have investigated the effect of neoclassical selfdiffusion in a realistic tokamak geometry with lower single null X-point. We also use a model with guiding center approximation in order to compare with the IMPGYRO full orbit model. The preliminary calculation results of each model have shown differences in the perpendicular average velocity of impurity ions at the top region of the SOL. The mechanism which leads to the difference has been discussed. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Status of the tokamak program

Science.gov (United States)

Sheffield, J.

1981-08-01

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

Soft x-ray imaging system for measurement of noncircular tokamak plasmas

International Nuclear Information System (INIS)

Fonck, R.J.; Reusch, M.; Jaehnig, K.P.; Hulse, R.; Roney, P.

1986-08-01

A soft x-ray camera and image processing system has been constructed to provide measurements of the internal shape of high temperature tokamak plasmas. The camera consists of a metallic-foil-filtered pinhole aperture and a microchannel plate image intensifier/convertor which produces a visible image for detection by a CCD TV camera. A wide-angle tangential view of the toroidal plasma allows a single compact camera to view the entire plasma cross section. With Be filters 12 to 50 μm thick, the signal from the microchannel plate is produced mostly by nickel L-line emissions which orignate in the hot plasma core. The measured toroidal image is numerically inverted to produce a cross-sectional soft x-ray image of the plasma. Since the internal magnetic flux surfaces are usually isothermal and the nickel emissivity depends strongly on the local electron temperature, the x-ray emission contours reflect the shape of the magnetic surfaces in the plasma interior. Initial results from the PBX tokamak experiment show clear differences in internal plasma shapes for circular and bean-shaped discharges

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

Integrated modeling of temperature profiles in L-mode tokamak discharges

Energy Technology Data Exchange (ETDEWEB)

Rafiq, T.; Kritz, A. H.; Tangri, V. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Voitsekhovitch, I. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Budny, R. V. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

2014-12-15

Simulations of doublet III-D, the joint European tokamak, and the tokamak fusion test reactor L-mode tokamak plasmas are carried out using the PTRANSP predictive integrated modeling code. The simulation and experimental temperature profiles are compared. The time evolved temperature profiles are computed utilizing the Multi-Mode anomalous transport model version 7.1 (MMM7.1) which includes transport associated with drift-resistive-inertial ballooning modes (the DRIBM model [T. Rafiq et al., Phys. Plasmas 17, 082511 (2010)]). The tokamak discharges considered involved a broad range of conditions including scans over gyroradius, ITER like current ramp-up, with and without neon impurity injection, collisionality, and low and high plasma current. The comparison of simulation and experimental temperature profiles for the discharges considered is shown for the radial range from the magnetic axis to the last closed flux surface. The regions where various modes in the Multi-Mode model contribute to transport are illustrated. In the simulations carried out using the MMM7.1 model it is found that: The drift-resistive-inertial ballooning modes contribute to the anomalous transport primarily near the edge of the plasma; transport associated with the ion temperature gradient and trapped electron modes contribute in the core region but decrease in the region of the plasma boundary; and neoclassical ion thermal transport contributes mainly near the center of the discharge.

Simulations of Tokamak Edge Turbulence Including Self-Consistent Zonal Flows

Science.gov (United States)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge is summarized in this mini-conference talk. A more detailed report on this work is presented in a poster at this conference. This work extends our previous work to include self-consistent zonal flows and their effects. The previous work addressed the simulation of L-mode tokamak edge turbulence using the turbulence code BOUT. The calculations used realistic single-null geometry and plasma parameters of the DIII-D tokamak and produced fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the US Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Physics design requirements for the Tokamak Physics Experiment (TPX)

International Nuclear Information System (INIS)

Neilson, G.H.; Goldston, R.J.; Jardin, S.C.; Reiersen, W.T.; Porkolab, M.; Ulrickson, M.

1993-01-01

The design of TPX is driven by physics requirements that follow from its mission. The tokamak and heating systems provide the performance and profile controls needed to study advanced steady state tokamak operating modes. The magnetic control systems provide substantial flexibility for the study of regimes with high beta and bootstrap current. The divertor is designed for high steady state power and particle exhaust

Two dimensional neutral transport analysis in tokamak plasma

International Nuclear Information System (INIS)

Shimizu, Katsuhiro; Azumi, Masafumi

1987-02-01

Neutral particle influences the particle and energy balance, and play an important role on sputtering impurity and the charge exchange loss of neutral beam injection. In order to study neutral particle behaviour including the effects of asymmetric source and divertor configuration, the two dimensional neutral transport code has been developed using the Monte-Carlo techniques. This code includes the calculation of the H α radiation intensity based on the collisional-radiation model. The particle confinement time of the joule heated plasma in JT-60 tokamak is evaluated by comparing the calculated H α radiation intensity with the experimental data. The effect of the equilibrium on the neutral density profile in high-β plasma is also investigated. (author)

Present status of Tokamak research

International Nuclear Information System (INIS)

Basu, Jayanta

1991-01-01

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

Tokamaks - Third Edition

International Nuclear Information System (INIS)

Rogister, A L

2004-01-01

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

Preliminary comparison of the conventional and quasi-snowflake divertor configurations with the 2D code EDGE2D/EIRENE in the FAST tokamak

Energy Technology Data Exchange (ETDEWEB)

Viola, B.; Maddaluno, G.; Pericoli Ridolfini, V. [EURATOM-ENEA Association, C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Rome) (Italy); Corrigan, G.; Harting, D. [Culham Centre of Fusion Energy, EURATOM-Association, Abingdon (United Kingdom); Mattia, M. [Dipartimento di Informatica, Sistemi e Produzione, Universita di Roma, Tor Vergata, Via del Politecnico, 00133 Roma (Italy); Zagorski, R. [Institute of Plasma Physics and Laser Microfusion-EURATOM Association, 01-497 Warsaw (Poland)

2014-06-15

The new magnetic configurations for tokamak divertors, snowflake and super-X, proposed to mitigate the problem of the power exhaust in reactors have clearly evidenced the need for an accurate and reliable modeling of the physics governing the interaction with the plates. The initial effort undertaken jointly by ENEA and IPPLM has been focused to exploit a simple and versatile modeling tool, namely the 2D TECXY code, to obtain preliminary comparison between the conventional and snowflake configurations for the proposed new device FAST that should realize an edge plasma with properties quite close to those of a reactor. The very interesting features found for the snowflake, namely a power load mitigation much larger than expected directly from the change of the magnetic topology, has further pushed us to check these results with the more sophisticated computational tool EDGE2D coupled with the neutral code module EIRENE. After a preparatory work that has been carried out in order to adapt this code combination to deal with non-conventional, single null equilibria and in particular with second order nulls in the poloidal field generated in the snowflake configuration, in this paper we describe the first activity to compare these codes and discuss the first results obtained for FAST. The outcome of these EDGE2D runs is in qualitative agreement with those of TECXY, confirming the potential benefit obtainable from a snowflake configuration. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

FIFPC, a fast ion Fokker--Planck code

International Nuclear Information System (INIS)

Fowler, R.H.; Callen, J.D.; Rome, J.A.; Smith, J.

1976-07-01

A computer code is described which solves the Fokker--Planck equation for the velocity space distribution of fast ions injected into a tokamak plasma. The numerical techniques are described and use of the code is outlined. The program is written in FORTRAN IV and is modularized in order to provide greater flexibility to the user. A program listing is provided and the results of sample cases are presented

Tokamak engineering mechanics

CERN Document Server

Song, Yuntao; Du, Shijun

2013-01-01

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

The tokamak as a neutron source

International Nuclear Information System (INIS)

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

1989-11-01

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

Commercial feasibility of fusion power based on the tokamak concept

International Nuclear Information System (INIS)

Reid, R.L.; Steiner, D.

1977-01-01

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

Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

International Nuclear Information System (INIS)

Xu, X.Q.; Cohen, R.H.; Porter, G.D.; Rognlien, T.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

2001-01-01

Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and nite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

International Nuclear Information System (INIS)

Xu, X.Q.; Cohen, R.H.; Por, G.D. ter; Rognlien, T.D.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

1999-01-01

Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the E x B drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

Dynamics and feedback control of plasma equilibrium position in a tokamak

International Nuclear Information System (INIS)

Burenko, O.

1983-01-01

A brief history of the beginnings of nuclear fusion research involving toroidal closed-system magnetic plasma containment is presented. A tokamak machine is defined mathematically for the purposes of plasma equilibrium position perturbation analysis. The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form. This symbolic form of the dynamics transfer function makes it possible to study the stability of a tokamak's plasma equilibrium position. Knowledge of the dynamics transfer function permits systematic syntheses of the required plasma displacement feedback control systems

GATO: An MHD stability code for axisymmetric plasmas with internal separatrices

International Nuclear Information System (INIS)

Bernard, L.C.; Helton, F.J.; Moore, R.W.

1981-01-01

The GATO code computes the growth rate of ideal magnetohydrodynamic instabilities in axisymmetric geometries with internal separatrices such as doublet and expanded spheromak. The basic method, which uses a variational principle and a Galerkin procedure to obtain a matrix eigenvalue problem, is common to the ERATO and PEST codes. A new coordinate system has been developed to handle the internal separatrix. Efficient algorithms have been developed to solve the matrix eigenvalue problem for matrices of rank as large as 40 000. Further improvement is expected using graph theoretical techniques to reorder the matrices. Using judicious mesh repartition, the marginal point can be determined with great precision. The code has been extensively used to optimize doublet and general tokamak plasmas. (orig.)

Conceptual design of a Langmuir probe system for the tokamak ASDEX-UPGRADE

International Nuclear Information System (INIS)

Anastassiadis, A.; Tsingas, A.C.; Tsois, N.N.; Zoumbos, G.A.

1985-05-01

The conceptual design of a Langmuir probe system for the tokamak ASDEX-UPG is presented. This system is intended to carry out electrostatic measurements, in space and time, on the boundary layer plasma over the largest possible volume of the divertor plasma during discharges. Conducted by preset design requirements a fast probe system is proposed. During discharges signal measurements will be performed by means of a data-acquisition system and the motion will be controlled by a real-time computer. The desired information concerning plasma parameters and the motion of the probe system will be available to the diagnostician via a video display unit. (author)

The design of the KSTAR tokamak

International Nuclear Information System (INIS)

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

1999-01-01

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

The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

International Nuclear Information System (INIS)

Lee, K. H.; Woo, H. K.; Im, K. H.; Cho, S. Y.; Kim, J. B.

2000-01-01

The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10 -6 ∼10 -7 Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses

The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

Energy Technology Data Exchange (ETDEWEB)

Lee, K. H.; Woo, H. K. [Chungnam National Univ., Taejon (Korea, Republic of); Im, K. H.; Cho, S. Y. [korea Basic Science Institute, Taejon (Korea, Republic of); Kim, J. B. [Hyundai Heavy Industries Co., Ltd., Ulsan (Korea, Republic of)

2000-07-01

The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6}{approx}10{sup -7}Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.

A low-cost ground loop detection system for Aditya-U Tokamak

International Nuclear Information System (INIS)

Kumar, Rohit; Kumawat, Devilal; Macwan, Tanmay; Ranjan, Vaibhav; Aich, Suman; Sathyanaryana, K.; Ghosh, J.; Tanna, R.L.

2017-01-01

Aditya-U is a medium sized Limiter-Divertor Tokamak machine. Different set of Magnetic Coils are installed for the generation of Magnetic field for the Plasma Initiation and Control in Pulse Mode. Support Structures with proper electrical Insulation are provided to Align and Hold these Magnetic Coils for the Plasma Operation. As machine operates at very high currents of kA’s range, very high vibrations are created during operations which can result in the breakdown of electrical insulation between different coils/systems/structures. The details of low cost ground loop detection system will be discussed in this paper

Design, simulation and construction of the Taban tokamak

Science.gov (United States)

H, R. MIRZAEI; R, AMROLLAHI

2018-04-01

This paper describes the design and construction of the Taban tokamak, which is located in Amirkabir University of Technology, Tehran, Iran. The Taban tokamak was designed for plasma investigation. The design, simulation and construction of essential parts of the Taban tokamak such as the toroidal field (TF) system, ohmic heating (OH) system and equilibrium field system and their power supplies are presented. For the Taban tokamak, the toroidal magnetic coil was designed to produce a maximum field of 0.7 T at R = 0.45 m. The power supply of the TF was a 130 kJ, 0–10 kV capacitor bank. Ripples of toroidal magnetic field at the plasma edge and plasma center are 0.2% and 0.014%, respectively. For the OH system with 3 kA current, the stray field in the plasma region is less than 40 G over 80% of the plasma volume. The power supply of the OH system consists of two stages, as follows. The fast bank stage is a 120 μF, 0–5 kV capacitor that produces 2.5 kA in 400 μs and the slow bank stage is 93 mF, 600 V that can produce a maximum of 3 kA. The equilibrium system can produce uniform magnetic field at plasma volume. This system’s power supply, like the OH system, consists of two stages, so that the fast bank stage is 500 μF, 800 V and the slow bank stage is 110 mF, 200 V.

Status of tokamak research

International Nuclear Information System (INIS)

Rawls, J.M.

1979-10-01

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

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

International Nuclear Information System (INIS)

Von Nessi, G T; Hole, M J

2014-01-01

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

Full Tokamak discharge simulation and kinetic plasma profile control for ITER

International Nuclear Information System (INIS)

Hee Kim, S.

2009-10-01

Understanding non-linearly coupled physics between plasma transport and free-boundary equilibrium evolution is essential to operating future tokamak devices, such as ITER and DEMO, in the advanced tokamak operation regimes. To study the non-linearly coupled physics, we need a simulation tool which can self-consistently calculate all the main plasma physics, taking the operational constraints into account. As the main part of this thesis work, we have developed a full tokamak discharge simulator by combining a non-linear free-boundary plasma equilibrium evolution code, DINA-CH, and an advanced transport modelling code, CRONOS. This tokamak discharge simulator has been used to study the feasibility of ITER operation scenarios and several specific issues related to ITER operation. In parallel, DINA-CH has been used to study free-boundary physics questions, such as the magnetic triggering of edge localized modes (ELMs) and plasma dynamic response to disturbances. One of the very challenging tasks in ITER, the active control of kinetic plasma profiles, has also been studied. In the part devoted to free-boundary tokamak discharge simulations, we have studied dynamic responses of the free-boundary plasma equilibrium to either external voltage perturbations or internal plasma disturbances using DINA-CH. Firstly, the opposite plasma behaviour observed in the magnetic triggering of ELMs between TCV and ASDEX Upgrade has been investigated. Both plasmas experience similar local flux surface expansions near the upper G-coil set and passive stabilization loop (PSL) when the ELMs are triggered, due to the presence of the PSLs located inside the vacuum vessel of ASDEX Upgrade. Secondly, plasma dynamic responses to strong disturbances anticipated in ITER are examined to study the capability of the feedback control system in rejecting the disturbances. Specified uncontrolled ELMs were controllable with the feedback control systems. However, the specifications for fast H-L mode

Burn Control Mechanisms in Tokamaks

Science.gov (United States)

Hill, M. A.; Stacey, W. M.

2015-11-01

Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

Axisymmetric instability in a noncircular tokamak: experiment and theory

International Nuclear Information System (INIS)

Lipschultz, B.; Prager, S.C.; Todd, A.M.M.; Delucia, J.

1979-09-01

The stability of dee, inverse-dee and square cross section plasmas to axisymmetric modes has been investigated experimentally in Tokapole II, a tokamak with a four-null poloidal divertor. Experimental results are closely compared with predictions of two numerical stability codes -- the PEST code (ideal MHD, linear stability) adapted to tokapole geometry and a code which follows the nonlinear evolution of shapes similar to tokapole equilibria. Experimentally, the square is vertically stable and both dee's unstable to a vertical nonrigid axisymmetric shift. The central magnetic axis displacement grows exponentially with a growth time approximately 10 3 poloidal Alfven times plasma time. Proper initial positioning of the plasma on the midplane allows passive feedback to nonlinearly restore vertical motion to a small stable oscillation. Experimental poloidal flux plots are produced directly from internal magnetic probe measurements

Influence of nonlinear effects on the neutral gas transport in tokamaks

International Nuclear Information System (INIS)

Behringer, T.

1992-06-01

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

Real-time control of Tokamak plasmas: from control of physics to physics-based control

International Nuclear Information System (INIS)

Felici, F. A. A.

2011-11-01

shown effective stabilization of both 3/2 and 2/1 NTMs, and have localized the most effective deposition location. Studies of current-profile driven destabilization of tearing modes in TCV plasmas with significant amounts of ECCD show a great sensitivity to details of the current profile, but failed to identify a stationary region in the parameter space in which NTMs are always destabilized, suggesting that transient effects play a role. The simultaneous control of magnetic and kinetic plasma profiles is another key requirement for advanced tokamak operation. While control of kinetic plasma profiles around an operating point can be handled using standard linear control techniques, the strongly nonlinear physics of the coupled profiles complicates the problem. Since internal magnetic quantities are difficult to measure with sufficient spatial and temporal resolution – even after years of diagnostic development – routine control of tokamak plasma profiles remains a daunting and challenging task. In this thesis, physics understanding of plasma current and energy transport is embedded in the control solution. The new lightweight transport code RAPTOR (RApid Plasma Transport simulatOR) has been derived focusing on simplicity and speed of simulation for real-time control. The partial differential equation for current diffusion is solved in real-time during a plasma shot in the TCV control system using RAPTOR. For the first time, this concept is applied experimentally to the tokamak current density profile problem. The real-time simulation gives a physics-model based estimate of key plasma quantities, to be controlled or monitored in real-time by different control systems. Any available diagnostics can be included into the simulation providing additional constraints and removing measurement uncertainties. The real-time simulation approach holds the advantage that knowledge of the plasma profiles is no longer restricted to those points in space and time where they are

Tokamak ARC damage

International Nuclear Information System (INIS)

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

1985-01-01

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

Tokamak ARC damage

Energy Technology Data Exchange (ETDEWEB)

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

1985-01-01

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

A three dimensional calculation of neutron streaming through ITER tokamak pumping ducts with the Monte Carlo code TRIPOLI-2

International Nuclear Information System (INIS)

Bresard, I.; Diop, C.M.; Giancarli, L.; Gervaise, F.

1991-01-01

In the frame of the ITER tokamak project, the streaming of neutrons through pumping ducts up to the properly so called pumping system is studied. The gas evacuation device of the ITER plasma consists of a set of vacuum pumps which are located in a room which is outside the main machine building. These pumps receive the exhaust gas through several pumping ducts with a cross section of about four square meters and a length of about ten meters. Although insensitive to the magnetic field, the 14 MeV neutrons from plasma D-T thermonuclear reactions can penetrate in the divertor and reach the room pumping device by propagation through the bent ducts. Different components of this system, such as the bellows, turbomolecular pumps, etc., are irradiated and that raises radiation problems. In this study we determine, by using 3D Monte Carlo transport code TRIPOLI-2, neutron fluxes, dose rates and heatings due to neutrons which have streamed out the plasma through the bent ducts, at several points of the pumping room. Results show the neutron flux attenuation reachs a factor 10 -5 from plasma chamber to the pumping hall; the neutron heatings are estimated to 1.9x10 -3 W/cm 3 in bellow stainless steel at duct entrance, and 8x10 -7 W/cm 3 in the turbopumping stainless steel structure, inside pumping hall. The neutron fluxes obtained will be used to compute gamma source produced by radiative, inelastic process and gamma rays from formed activation products. Then, the knowledge of gamma source will allow to compute gamma dose rate and heating. The dose rates and heatings obtained will contribute to the definition of the ITER pumping system technical options and to establish pumping hall access conditions, also. (orig.)

Plasma driving system requirements for commercial tokamak fusion reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

1978-01-01

The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results presented in this paper can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

Plasma driving system requirements for commercial tokamak fusion reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

1977-01-01

The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

Equilibrium system analysis in a tokamak ignition experiment

International Nuclear Information System (INIS)

Carrera, R.; Weldon, W.F.; Woodson, H.H.

1989-10-01

The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies

Equilibrium system analysis in a tokamak ignition experiment

Energy Technology Data Exchange (ETDEWEB)

Carrera, R.; Weldon, W.F.; Woodson, H.H.

1989-10-01

The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

Plasma density remote control system of experimental advanced superconductive tokamak

International Nuclear Information System (INIS)

Zhang Mingxin; Luo Jiarong; Li Guiming; Wang Hua; Zhao Dazheng; Xu Congdong

2007-01-01

In Tokamak experiments, experimental data and information on the density control are stored in the local computer system. Therefore, the researchers have to be in the control room for getting the data. Plasma Density Remote Control System (DRCS), which is implemented by encapsulating the business logic on the client in the B/S module, conducts the complicated science computation and realizes the synchronization with the experimental process on the client. At the same time, Web Services and Data File Services are deployed for the data exchange. It is proved in the experiments that DRCS not only meets the requirements for the remote control, but also shows an enhanced capability on the data transmission. (authors)

BIRTH: a beam deposition code for non-circular tokamak plasmas

International Nuclear Information System (INIS)

Otsuka, Michio; Nagami, Masayuki; Matsuda, Toshiaki

1982-09-01

A new beam deposition code has been developed which is capable of calculating fast ion deposition profiles including the orbit correction. The code incorporates any injection geometry and a non-circular cross section plasma with a variable elongation and an outward shift of the magnetic flux surface. Typical cpu time on a DEC-10 computer is 10 - 20 seconds and 5 - 10 seconds with and without the orbit correction, respectively. This is shorter by an order of magnitude than that of other codes, e.g., Monte Carlo codes. The power deposition profile calculated by this code is in good agreement with that calculated by a Monte Carlo code. (author)