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A novel hybrid method for accurate magnetic field measurements in steady-state or long-pulse tokamaks, described in a 1995 publication and patented in 1998, has not yet been tested in a tokamak. Recent development progress is described herein.

The primary goal of the FIRE preconceptual design is to affordably conduct physics experiments in the fusion ignition regime with self-heated plasmas. The device could also permit advanced tokamak experiments in deuterium lasting several minutes. Tradeoff studies considering MVA and flattop duration have identified TF and PF magnet power system designs expected to have low cost, consistent both with the mission to enter the ignition regime and the possibility to conduct long pulse deuterium experiments. In addition, a performance-extending upgrade path for the power system has been identified which could be followed later, if experimental results justify further increasing the maximum toroidal field and plasma current or lengthening the duration.

The authors describe the methods used at PPPL (Princeton Plasma Physics Laboratory) to: (1) locate and characterize multiple electrical leakage paths in the insulation of the TFTR (Tokamak Fusion Test Reactor) toroidal field (TF) coils; and (2) determine if turn to turn shorts exist in the coils. The electrical leakage test was first used in May of 1991 to locate a single leakage path of 15 M Omega in TF coil No. 3. The test was again used to locate a single leakage path of 1 M Omega in TF coil No. 18 on September 3, 1991. In theory, leakages as high as 75 M Omega can be located and characterized using this method. The turn to turn test consists of individual tests on each TF coil and a signature test to determine if changes have taken place since the last test. The turn to turn portion of the test was performed on the TFTR TF coils in April of 1991. The signature test was performed in July of 1991. Tests have shown that turn to turn shorts as high as 40 m Omega (500 times the 80 mu Omega resistance of a single turn) can be located using this method. The hardware designed and built to support both tests and the test results are also described. >

Engineering issues relevant to the design of magnetic based systems for plasma feedback stabilization of internal tearing modes and resistive wall modes are discussed herein. Proposed design optimization methods are delineated, and are then illustrated in practice by applying them to the example design of a hypothetical experimental plasma feedback stabilization demonstration facility. In addition to the physical dynamics of the plasma itself and of the eddy current behavior of metallic structures close to the plasma, the plasma feedback system includes magnetic field sensors distributed spatially near the plasma, digital signal processing electronics to implement an optimal state-estimating "observer" and an optimal controller, and also electromagnets and their associated power circuitry. The magnetic field sensors must be sufficiently numerous and spatially well distributed so that each unstable plasma eigenmode can be resolved without spatial aliasing. Digital signal processing algorithms must be sufficiently fast to continually estimate the amplitude, phase, and rotation rate of each unstable eigenmode based on the magnetic field sensors' signal histories. They must be sufficiently sophisticated to discriminate between field amplitude components produced directly by the plasma perturbations and the components produced in response to optimal controller commands, and they should be sufficiently robust for a range of plasma parameters. The electromagnets must be located close enough to the plasma to drive a controlled opposing field with the proper resolution of helical shapes, and their power circuitry must be capable of producing the commanded amplitude, phase, and rotation rates for the driven field.

A new hybrid method and apparatus is presented for poloidal magnetic field measurement, suitable for use in steady-state control of tokamak plasma shape, position, and current. [An invention disclosure has been filed.] It combines two different magnetic field principles (induction and torque) into a single hybrid measurement apparatus, thus in one device providing accurate magnetic field measurement over the entire frequency spectrum from DC to several kilohertz.

An alternate cooling fluid (Fluorinert/sup TM/) was introduced during the D-T experimental runs for cooling the toroidal field (TF) coils on TFTR. This paper addresses how this system performed during the D-T operational period and the special techniques that the alternate cooling system requires for operations and maintenance. Radiation from neutron activation of Fluorinert/sup TM/ (fluorine 18 etc) and associated personnel safety issues and safeguards are discussed. Flow reversal in the TF Coils has proven to be a valuable mechanism by which any loose particulates can be cleared from the coolant passages and the development of this operational tool is addressed. Specific draining procedures for the TF Coils have been developed during the D-T run in order to minimize losses of fluid. The toroidal field coil operational characteristics with Fluorinert/sup TM/ are compared to those previously observed with water cooling.

For the past year, TFTR has sometimes operated at extended toroidal field (TF) levels, The extension to 5.6 Tesla (79 kA) was crucial for TFTR's November '94 10.7 MW DT fusion power record. The extension to 6.0 Tesla (85 kA) was commissioned on 9 September 1995. There are several reasons that one could expect the TF coils to survive the higher stresses that develop at higher fields. They were designed to operate at 5.2 Tesla with a vertical field of 0.5 Tesla, whereas the actual vertical field needed for the plasma does not exceed 0.35 Tesla. Their design specification explicitly required they survive some pulses at 6.0 Tesla. TF coil mechanical analysis computer models available during coil design were crude, leading to conservative design. And design analyses also had to consider worst-case misoperations that TFTR's real time Coil Protection Calculators (CPCs) now positively prevent from occurring. Engineering considerations are summarized.

Liquid metal walls have the potential solve to first-wall problems for fusion reactors, such as heat load and erosion of dry walls, neutron damage and activation, and tritium inventory and breeding. In the near term, such walls can serve as the basis for schemes to stabilize magnetohydrodynamic (MHD) modes. Furthermore, the low recycling characteristics of lithium walls can be used for particle control. Liquid lithium experiments have already begun in the Current Drive eXperiment-Upgrade (CDX-U). Plasmas limited with a toroidally localized limiter have been investigated, and experiments with a fully toroidal lithium limiter are in progress. A liquid surface module (LSM) has been proposed for the National Spherical Torus Experiment (NSTX). In this larger ST, plasma currents are in excess of 1 MA and a typical discharge radius is about 68 cm. The primary motivation for the LSM is particle control, and options for mounting it on the horizontal midplane or in the divertor region are under consideration. A key consideration is the magnitude of the eddy currents at the location of a liquid lithium surface. During plasma start up and disruptions, the force due to such currents and the magnetic field can force a conducting liquid off of the surface behind it. The Tokamak Simulation Code (TSC) has been used to estimate the magnitude of this effect. This program is a two dimensional, time dependent, free boundary simulation code that solves the MHD equations for an axisymmetric toroidal plasma. From calculations that match actual ST equilibria, the eddy current densities can be determined at the locations of the liquid lithium. Initial results have shown that the effects could be significant, and ways of explicitly treating toroidally local structures are under investigation.

The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. Sputtering and erosion tests are currently underway in the PISCES device at the University of California at San Diego (UCSD). To complement this effort, plasma interaction questions in a toroidal plasma geometry will be addressed by a proposed new groundbreaking experiment in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The CDX-U plasma is intensely heated and well diagnosed, and an extensive liquid lithium plasma-facing surface will be used for the first time with a toroidal plasma. Since CDX-U is a small ST, only approximately1 liter or less of lithium is required to produce a toroidal liquid lithium limiter target, leading to a quick and cost-effective experiment.

Fusion energy is a potential energy source for the future with plentiful fuel supplies and is expected to have benign environmental impact. The issue with fusion energy has been the scientific feasibility, and recently the cost of this approach. The key technical milestone for fusion is the achievement of a self-sustained fusion fire, ignition, in the laboratory. Despite 40 years of research and the expenditure of almost $20B worldwide, a self-sustained fusion fire has not yet been produced in the laboratory. The fusion program needs a test bed, preferably more than one, where the dynamics of a burning plasma can be studied, optimized and understood so that the engineering requirements for an engineering test reactor can be determined. Engineering and physics concepts must be developed within the next decade that will lead to an affordable burning plasma experiment if fusion is going to be perceived as making progress toward a potential long range energy source.

Early in 1987, the TFTR's magnetic field system will be upgraded. The planned improvements concentrate on improving the ability to inductively drive plasma currents, both by increasing the number of turns in the Ohmic Heating (OH) coil system through an external reconnection of existing coils, and by increasing the OH operating current swing. The most novel aspect of this upgrade is the new Coil Protection Calculator, which will allow operation of coils up to the maximum levels possible without exceeding coil stress limits. Other changes include increases in OH coil power supply current, voltage, and energy capabilities.