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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. >

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.

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.