Adiabatic Compression of a Doublet Field Reversed Configuration (FRC)

The adiabatic compression of magnetized plasmas has come to the fore in recent times as an interesting hybrid of both inertial and magnetic fusion energy schemes, possibly allowing a means to reach fusion conditions in a compact pulsed system (R.P. Drake et al. Fusion Tech. 30, 310, (1996)). It is possible to compress a range of different magnetic configurations (D.D. Ryutov, R.E. Siemon, Com. Mod. Phys. 2, 185, (2001)), here we consider the compression of a FRC: a favorable target due to high ß. The literature relating to the adiabatic compression of magnetic concepts is reviewed. We present analytic modeling and MHD simulations of the reconnection and compression of a doublet FRC configuration that might serve as a target for compression.     

Axisymmetric Toroidal Equilibrium with Sheared Toroidal Flows

Problem of the axisymmetric toroidal equilibrium with pure sheared toroidal flow is involved.For standard tokamak equilibrium,general approximate solutions are analytically pursued for arbitrary current profile and non-circular cross-section.Equilibrium properties including the flow-induced density asymmetry are analyzed.     

Analytical Technique for Determination of Toroidal Plasma Displacement

An analytical technique for the determination of toroidal plasma displacement is presented. First, the plasma horizontal position is calculated from the external vertical field coil characteristics. The calculation is made focusing on the external vertical field coil current and voltage changes due to a horizontal displacement of plasma column. Also for comparison of result, a set of magnetic probes were designed, constructed and used. The results from these two techniques are compared and discussed.     

Toroidal Field Ripple reduction studies for ITER and FAST

Two different approaches to control the Toroidal Field Ripple (TFR) amplitude in ITER and FAST devices are presented in this paper. The approach currently adopted to reduce the TFR in ITER is based on the installation of ferromagnetic inserts between the vacuum vessel shells. The same approach has been analyzed in the design of the Fusion Advanced Studies Torus (FAST) proposal. Details of the system's layout are given. A new approach based on the insertion of active coils between the outer legs of the Toroidal Field Coils (TFCs) and the plasma, has been extensively investigated for these two machines. This active system would allow reducing the TFR to values even smaller than with the ferromagnetic inserts. The case of a localized disturb like that introduced by a Test Blanket Module (TBM) for ITER is presented where only well localized active coils can produce a significant ripple reduction.     

A Study on Main Breaker for Quench Protection of HT-7U Toroidal Superconducting Tokamak

This paper proposes a quench protection project of HT-7U toroidal superconductingtokamak through a forced commutation analysis of DC circuit breaker (DCCB) paralleling fuse.Based on the requirement of quench protection, main parameters are selected. Experimentalresults demonstrate the validity of this proposed project.     

Integrated Design System of Toroidal Field Coil for CFETR

Integrating engineering software is meaningful but challenging for a system code of a fusion device.This issue is seldom considered by system codes currently.Therefore,to discuss the issue,the Integrated Design System of TF Coil(IDS-TFC) has been worked out,which consists of physical calculation,CAD,and Finite Element Analysis(FEA).Furthermore,an Integrated and Automatically Optimized Method(IAOM) has been created to address the integration and interfaces.The method utilizes a geometry parameter to connect each design submodule and achieve automatic optimization.Double-objectives optimization has been realized,confirming it is feasible to integrate and optimize engineering design and physical calculation.Moreover,IDSTFC can also serve as a useful reference of integrated design processing for subsequent fusion design.     

Design of a Wedge-Shaped Toroidal Field Winding for KTX Device

The KTX device is a reversed field pinch(RFP)device currently under construction.Its maximum plasma current is designed as 1 MA with a discharge time longer than 100 ms.Its major radius is 1.4 m and its minor radius is 0.55 m.One of the most important problems in the magnet system design is how to reduce the TF magnetic field ripple and error field.A new wedgeshaped TF coil is put forward for the KTX device and its electromagnetic properties are compared with those of rectangular-shaped TF coils.The error field Bn/Btof wedge-shaped TF coils with6.4 degrees is about 6%as compared with 8%in the case of a rectangular-shaped TF coil.Besides,the wedge-shaped TF coils have a lower magnetic field ripple at the edge of the plasma region,which is smaller than 7.5%at R=1.83 m and 2%at R=1.07 m.This means that the tokamak operation mode may be feasible for this device when the plasma area becomes smaller,because the maximum ripple in the plasma area of the tokamak model is always required to be smaller than 0.4%.Detailed analysis of the results shows that the structure of the wedged-shape TF coil is reliable.It can serve as a reference for TF coil design of small aspect ratio RFPs or similar torus devices.     

The Toroidal Magnet of FM-1

The construction features of the toroidal field coil for the FM-1 device are described. The coil is rated at 1.5 × 106 ampere-turns steady and has a field volume of 49 cubic meters. The stored energy of the field is nearly 2 MJ. The coil must be taken apart for insertion of a superconducting ring; the techniques for accomplishing this are described.     

Point-Source Helicity Injection Current Drive System for the Pegasus Toroidal Experiment

Recent Pegasus experiments are developing solenoid-free startup techniques using point-source magnetic helicity injection. These plasma sources, called “plasma guns”, ionize a stream of gas in a discharge channel, and bias this channel with respect to an external electrode, driving current along the plasma stream, which relaxes into a tokamak-like equilibrium. The relaxed discharges formed by these injectors exhibit high current amplification, which is the ratio of total toroidal current to the gun-driven current. The development and present design of these injectors are described, and time traces from a typical discharge are presented.     

Advanced Power Conversion Efficiency in Inventive Plasma for Hybrid Toroidal Reactor

Apex hybrid reactor has a good potential to utilize uranium and thorium fuels in the future. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together. The most important feature of hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity good power transformation productivity, the magnitude of the reactor’s operational duration, low failure percentage, short maintenance time and the inclusion of the system’s simple technology and material. The analysis has been made using the MCNP Monte Carlo code and ENDF/B–V–VI nuclear data. Around the fusion chamber, molten salts Flibe (LI₂BeF₄), lead–lithium (PbLi), Li–Sn, thin-lityum (Li₂₀Sn₈₀) have used as cooling materials. APEX reactor has modeled in the torus form by adding nuclear materials of low significance in the specified percentages between 0 and 12 % to the molten salts. In this study, the neutronic performance of the APEX fusion reactor using various molten salts has been investigated. The nuclear parameters of Apex reactor has been searched for Flibe (LI₂BeF₄) and Li–Sn, for blanket layers. In case of usage of the Flibe (LI₂BeF₄), PbLi, and thin-lityum (Li₂₀Sn₈₀) salt solutions at APEX toroidal reactors, fissile material production per source neutron, tritium production speed, total fission rate, energy reproduction factor has been calculated, the results obtained for both salt solutions are compared.     

A Consideration on Increasing Current Density in Normal Conducting Toroidal Field Coil for Spherical Tokamak Power Plant

The center post is the most critical component as an inboard part of the toroidal field coil for the low aspect ratio tokamak. During the discharge it endures not only a tremendous ohmic heating owing to its carrying a rather high current but also a large nuclear heating and irradiation owing to the plasma operation. All the severe operating conditions, including the structure stress intensity and the stability of the structure, largely limit the maximum allowable current density. But in order to contain a very high dense plasma, it is hoped that the fusion power plant system can operate with a much high maximum magnetic field BT ≥12 T-15 T in the center post. A new method is presented in this paper to improve the maximum magnetic field up to 17 T and to investigate the possibility of the normal conducting center post to be used in the future fusion tokamak power plant.     

A 57-Pin Langmuir Probe Array for Two-Dimensional Diagnosis in the Toroidal KT-5D Steady-State Plasma

In this article, the design idea of a 57-pin Langmuir probe array is presented, together with some experimental results. The manufacture technique of the probe array is discussed. Two-dimensional distributions of the plasma density, the floating potential and their fluctuations in the poloidal cross section are obtained. The blobs, which move radially and poloidally, cause great particle and energy transport. The charge neutralization by the poloidal movement of particles greatly decreases the vertical electric field in the helimak plasma. This leads to a confinement time which is two orders of magnitude larger than the theoretical prediction.     

An Adiabatic Model for Space-Dependent Boiling Water Reactor Kinetics

This and the companion paper⁽¹⁾ deal with space-dependent boiling water reactor kinetics by the method of adiabatic approximation. In this paper, equations are derived for axially space-dependent BWR feedback kinetics. These equations are coupled to the neutron equations through adiabatic approximation.

The reactor is divided into a number of axial regions. Assuming a trapezoidal distribution of the void and the water velocity distributions along the axis, balance equations for energy and mass are integrated to yield the region-exit quantities. With other pertinent equations, a computer code is written.

Some numerical results for a sample BWR are presented in the companion paper.     

Toroidal Multipolar Expansion for Fast L-Mode Plasma Boundary Reconstruction in EAST

A new method for plasma boundary reconstruction, based on the toroidal multipolar expansion (TME) scheme, is applied successfully in EAST. TME applies a limited number of toroidal multipolar moments based on toroidal coordinates to treat a two-dimensional problem of axisymmetric plasma equilibrium. The plasma boundary reconstructed by TME is consistent with the results by using EFIT. The method is su±ciently reliable and fast for real time shape control.     

Four-fluid Model and Radial Transport Equation for Toroidal Momentum in Tokamak Plasmas

1. IntroductionIn order to explain some recent experiment resultsand find new ways to improve the plasma performance: the study of the toroidal momentum transport and the spacial structure of radial electric fieldhave drawn more and more attention in tokamakplasma research [l,2,3,4]. To date, the theoreticalanalysis of plajsma toroidal momentum transport hasbeen mainly based on the ordinary MHD model.However, the ordinary MHD model fails to deal withconstraint conditions on the motions of th…     

Toroidal High-Density Singularity in a Small Plasma Focus

A toroidal singularity in the current sheet of a Plasma Focus operated at tens of joules (PF-50J) leading to the formation of a high-density singularity on the plasma sheath was observed on interferographic images of discharges in 6 mbar of Deuterium gas. The phenomenon was numerically simulated by means of the 2D numerical code Cshock, which was able to reproduce the plasma kinetics and the evolution of the electrical variables, as well as the formation, movement and density of the plasmoid structure.     

The Analysis and Calculation for the Toroidal Magnetic Field of HT-7U

1. IlltroductionThe HT--7U device is a large noncircular cross section super--conducting tokamak to be built in 2003.It will have a long pulse (60 ~ 1000 s) capability, aflexible PF system, and auxiliary heating and current driving sXstemsl and will be able to accommodate divertor heat loads that make 'it an attractivetest for the development of advanced tokamak operating modes [l]. The overview structure of theHT--7U tokamak is shown in Fig.1. Its toroidalfield system is one of important …     

Study of Electron Orbits for Formation of Toroidal Closed Flux Surface by ECH

Electron orbits under the external vertical field (BV) and the self poloidal field of the toroidal plasma current in a toroidal geometry are investigated by using analytic model fields in order to search for the conditions of occurence of appropriate confinement asymmetry of fast electrons along the field lines. This asymmetry e±ciently and quickly generates a toroidal current, which may close the field lines in ECH-started plasmas. It is analytically shown that the characteristics of confinement asymmetry depend on the product of the major radius and the strength of the external vertical field. The results suggest that in large devices BV should be lowered and also some artificial means to speed up the pitch angle scattering for the fast electrons at a higher energy range are beneficial to generate the toroidal current in order to close the filed lines.