New IQ demodulator, which will be used in the reflectometry system, was developed for the Alcator C-Mod experiments. Because the demodulator itself is a 4 mm × 4 mm integrated circuit chip and has very wide operational bandwidth (0.7 GHz - GHz), it can be mounted on a circuit board with its low frequency IQ outputs connecting directly to gain control and dc offset adjustment circuits. This whole circuitry is realized on a 4"×2.5" printed circuit board and can be well tailored to meet the requirements for any type of data collection system. Since the RF and LO ports are operated at very narrow bandwidth in this application, single stub match circuits are designed to improve the ports matching as well as serve as a single pole filter.
Coaxial Helicity Injection (CHI), a non-inductive method to initiate plasma and generate toroidal plasma current, is being investigated in the National Spherical Torus Experiment (NSTX). The center stack and outer vacuum vessel are separated by insulating gaps at the top and bottom of the slim central column so that a high voltage (up to 2 kV) can be applied between them from a pulsed power supply or a capacitor bank to initiate an arc discharge. In the presence of a suitable poloidal magnetic field, the discharge is initiated at the lower gap (the injector gap) and because of the strong toroidal field develops a helical structure resulting in substantial toroidal plasma current being driven. In NSTX, up to 390 kA of toroidal current has been generated for an injected current of 25 kA. The early investigations of CHI however frequently developed arcs across the insulator at the top of the machine (the absorber gap), which terminated the desired discharge. This arcing greatly restricted the operational space available for CHI studies. During 2002, the absorber region was modified to suppress these arcs. The new design includes a new ceramic insulator on the high field side of the absorber region with a much longer tracking distance between conducting elements at the different potentials. Furthermore, two new coils were installed near the absorber to provide the ability to minimize the poloidal field connecting the center stack and outer vacuum vessel. During the subsequent experimental campaign, CHI operation was less prone to arcing in the absorber and those arcs that did occur did not terminate the main discharge
Many advances have been made to the DIII-D plasma control simulation environment since the previously developed hardware-in-the-loop plasma shape simulation capability was reported. In the present paper we summarize the major improvements to this simulation environment, including, introduction of the non-linear plasma evolution code DINA. Comparisons with DIII-D experimental results are presented. Recent model developments in advanced neoclassical tearing mode (NTM) and resistive wall mode (RWM) control are presented.
The JET-EP antenna is designed to verify the design principles of that proposed for ITER, in relevant plasma conditions. It is intended to launch 7.2 MW of RF power to the plasma of the JET machine during its 'Enhanced Performance' program, at a power density of around 8 MW/m/sup 2/. The antenna comprises two poloidal current straps, each subdivided into four electrically short straps complete with in-vessel capacitors. The antenna is supported via a cantilever support box to the external support structure. This plug-in, through-port type assembly is clamped to the vacuum vessel externally. The antenna itself is fitted remotely. External rails and bellows allow radial positioning of the assembly. Eight capacitors are situated just behind the short antenna current straps. Plug-in replacement of the capacitors is performed through the vacuum vessel port.
The modification of JT-60 is planned as a fully superconducting tokamak (JT-60SC). The mission of JT-60SC program is to establish scientific and technological bases for an advanced operation in an economically and environmentally attractive DEMO reactor and ITER. The research objectives are to accomplish high performance steady state operation with high beta and non-inductive full current drive, with high bootstrap current fraction, and demonstrate the plasma applicability of reduced-activation material for a plasma of break-even class relevant to the reactor plasma. Basic design of JT-60SC has been completed and the detailed design is under way. The engineering design for main components of JT-60SC is described.
SiC (silicon carbide) has been considered as one of the primary candidate materials for a first wall component in a future fusion reactor because it has been claimed that SiC has excellent high temperature properties and low activation. However, the behavior of tritium on SiC has not been discussed yet. In this study, the tritium trapping capacity on the surface of SiC is experimentally obtained as about 10/sup 10/ Bq/m/sup 2/ at the temperature range of 298K- 1073K in consideration of the static system effect. It is also observed that tritium is trapped to the surface through hydrogen isotope exchange reaction. The isotope exchange reaction rate between tritiated water in a gas phase and hydrogen on the surface is quantified at the temperature of 298K, 773K, 973K in consideration of the kinetic system effect by the numerical curve fitting method applying the serial reactor model. The reaction rate is observed to be constant at 3.48 /spl times/ 10/sup -5/ m/s. The comparison of tritium properties of SiC with that of graphite or stainless steel is also discussed.
The proposal is to make large deuterium (D-D) magnetic fusion power plants in which some (most) of the tritium produced by fusion is removed and stored. This tritium will ultimately decay to helium-3 that will be recycled to supplement the helium-3 produced by fusion. Thus the dominant fusion becomes that of deuterium and helium-3. The level of neutron damage is reduced very substantially from that for a D-T power plant.
This paper presents initial results of a task to identify accident initiating events for inertial fusion energy (IFE) power plant designs. Initiating events (IEs) are a fundamental building block of a probabilistic risk assessment; they are the 'accident starters' that are analyzed to determine the risks posed to members of the public in the vicinity of the power plant. The IE results for the SOMBRERO design are presented in tabular form. The SOMBRERO design was analyzed since it is representative of dry chamber wall, laser driven designs. This work is used to characterize IFE plant risk and to identify potential design changes that would mitigate the plant risk.
Magnetohydrodynamic (MHD) stability analyses of Negative Central Shear (NCS) equilibria have revealed a new understanding of the limiting MHD instabilities in NCS experiments. Ideal stability calculations show a synergistic effect between cross-section shape and pressure profile optimization; strong shaping and a broader pressure profile independently lead to moderately higher beta limits, but broadening of the pressure profile in a strongly dee-shaped cross-section leads to a dramatic increase in the ideal beta limit. Localized resistive interchange (RI) modes can be unstable in the negative shear region and are most restrictive for peaked pressure profiles. Resistive global modes can also be destabilized significantly below the ideal beta limit. Experiments largely confirm the general trends, and diagnostic measurements and numerical stability calculations are found to be in good qualitative agreement. Observed disruptions in NCS discharges with L mode edge and strongly peaked pressure appear to be initiated by interactions between the RI and the global ideal and resistive modes.
Through time dependent transport simulation, we have developed an operation scenario with high bootstrap, negative shear configuration for KT-2 tokamak, so that investigations of the so-called "advanced tokamak operations" are possible. In this study, we have concentrated on calculation of power requirements of heating and current drive system, MHD (Magnetohydrodynamic) stability and compatibility with PF (Poloidal Field) coil system. The study shows that high bootstrap (>70%), MHD stable (up to /spl beta//sub N//spl les/3.9) operation scenario with negative shear configuration is possible within KT-2 tokamak design specification.
The paper discusses some aspects of the safety analysis currently being performed for a compact high-field tokamak device (IGNITOR). The analysis has shown that the radioactive releases from the last containment barrier that could occur during major accidents are negligible when comparing IGNITOR with similar fusion devices. The paper focuses on some of the design solutions, suggested by the analysis, in order to improve the overall plant availability.
The design of the cathode grid for a spherical inertial electrostatic confinement (IEC) fusion device is described based upon the results of SIMION, an electric field and ion trajectory program. Cathode grids of varying radii and geometric transparency were simulated with this code. Two-dimensional models of energetic ions in the IEC that were produced with SIMION show the formation of ion microchannels and the Star-mode. The SIMION results also show that smaller, less transparent grids produce smaller cores and more focused ion channels
Inertial electrostatic confinement (IEC) is one of the earliest plasma confinement concepts, having first been suggested by P.T. Farnsworth in the 1950s. The concept involves a simple apparatus of concentric spherical electrostatic grids or a combination of grids and magnetic fields. An electrostatic structure is formed from the confluence of electron or ion beams. Gridded IEC systems have demonstrated neutron yields as high as 2*10 neutrons/sec in steady state. These systems have considerable potential as small, inexpensive, portable neutron sources for assaying applications. Neutron tomography is also a potential application. This paper discusses the IEC concept and how it can be adapted to a steady-state assaying source and an intense pulsed neutron source. Theoretical modeling and experimental results are presented
Neutral beam injection has been the most successful scheme used to heat magnetically confined plasmas studied in controlled nuclear fusion research, and neutral beams are a candidate to heat to ignition the International Tokamak Experimental Reactor (ITER). This article describes the system which is presently being designed in Europe, Japan, and Russia, with coordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D0 to the ITER plasma for pulse length of ≳1000 s. The proposed injectors each use a single caesiated volume arc discharge negative ion source, and a multigrid, multiaperture accelerator, to produce about 40 A of 1 MeV D−. This will be neutralized in a subdivided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped onto the water-cooled surfaces making up the electrostatic residual ion dump. A water-cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER.
The increased use of UNIX based computer systems for machine control, data handling and analysis has greatly enhanced the operating scenarios and operating efficiency of the DIII-D tokamak. This paper will describe some of these UNIX systems and their specific uses. These include the plasma control system, the electron cyclotron heating control system, the analysis of electron temperature and density measurements and the general data acquisition system (which is collecting over 130 Mbytes of data). The speed and total capability of these systems has dramatically affected the ability to operate DIII-D. The improved operating scenarios include better plasma shape control due to the more thorough MHD calculations done between shots and the new ability to see the time dependence of profile data as it relates across different spatial locations in the tokamak. Other analysis which engenders improved operating abilities will be described.
The paper describes the “electronic” stiffening of the actuators developed for the articulated boom of JET. Experimental trials on a two joints boom have confirmed the improvements of speed and reliability that can be achieved. Based on these developments, a new control system for the boom is now under development together with a simulator intended to make easier and faster the system commissioning, design and optimization as well as the fault identification. The main originalities of the paper are the modeling of the actuators, based on a mechanical equivalent and an extension of the Craig's computed torque method to elastic systems
The first injector of the ASDEX Upgrade neutral beam system has been commissioned and injection experiments started in June 1993. Within a few weeks an injected power of 5.2 MW (H0), equivalent to 87% of the design power level, was achieved for pulse lengths of up to 1.5 seconds, and a variety of neutral beam heated discharges could be analyzed before the shutdown of ASDEX Upgrade in August
