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Calorimeter for pulsed energy sources

Description: A common problem in plasma physics experiments is the measurement of the energy deposited on a water-cooled plate by a pulsed energy source. Examples of this are neutral-beam-line defining plate and targets and tokamak water-cooled limiters. One method of measuring this energy is to integrate the product of the temperature rise (..delta..T) and the flow rate (F) of the coolant over the interval between the pulses. The two input parameters ..delta..T and F are derived from a differential thermopile and a turbine flow meter, respectively. A simple digital readout circuit displays the deposited energy in a light-emitting diode display. The circuit uses a commercially available, dual-slope analog-to-digital converter (ADC) in a novel configuration that multiplies ..delta..T and F directly. Calibration of the readout circuit is quick and simple, and short-term accuracies of 5% are easily obtained. Over longer periods the accuracy becomes degraded, primarily by thermal drifts in the thermopile amplifier and in the thermopile connections and the wiring. This offset must be compensated for by a simple adjustment before each experimental run. This readout circuit has been used successfully on the Impurity Study Experiment (ISX-B) neutral beam lines and is most convenient for deposited energy measurements in applications in which computerized data acquisition is not available.
Date: January 1, 1979
Creator: Pearce, J.W. & Edmonds, P.H.
Partner: UNT Libraries Government Documents Department

A calorimeter for measuring the neutral beam power reaching the plasma

Description: A calorimeter has been designed to measure the neutral beam power reaching the Advanced Toroidal Facility (ATF) plasma. The high-heat-flux surface of this calorimeter is made of an array of graphite tiles. The calorimeter, which will be located in the adapter section between the ATF vacuum vessel and the beam line, is retractable so that it can be moved away from the plasma without opening the vacuum vessel during normal ATF operation. Two rows of thermocouples mounted perpendicular to each other allow determination of the beam profile. This paper presents the details of the design and fabrication of the calorimeter. 5 refs., 6 figs.
Date: January 1, 1987
Creator: Menon, M.M.; Edmonds, P.H. & Hahs, C.L.
Partner: UNT Libraries Government Documents Department

High beta results in ISX-B with intense neutral beam injection

Description: Experiments on the ISX-B device show a deterioration in confinement at high beam power. In particular the electron energy confinement time falls catastrophically with increasing beam power. The maximum volume averaged beta values achieved are <2.5%; this is much less than would be predicted by extrapolating the low power data. Elongation has not been observed to have any significant effect on the maximum attainable beta, perhaps due to the limited range of both internal and external elongation. The electron energy confinement time does not follow Alcator scaling at high injection powers. There are two likely candidates for the loss of confinement. The phenomena may be ..beta../sub p/ specific and caused by the gradual onset of resistive MHD pressure driven modes producing deteriorating confinement through fluctuations in the poloidal magnetic field. Alternatively the phenomena may be specific to the method of heating, neutral injection, being caused, for example, by plasma rotation, where the rotation speed approaches the ion thermal velocity. Experiments are in progress to investigate both of these possibilities.
Date: January 1, 1982
Creator: Edmonds, P.H.; Bates, S.C. & Bell, J.D.
Partner: UNT Libraries Government Documents Department

Impurity behavior in the ISX-B tokamak

Description: Any discussion of impurity behavior during neutral-beam injection in ISX-B is best formulated in terms of the distinctive differences observed between ohmically heated, co-injection, and counter-injection discharges. In ohmically heated discharges both the production and the transport of impurities depend upon whether the working gas is hydrogen or deuterium. The influx of oxygen is almost the same for both cases, but the influx of metals is about a factor of 3 larger in the deuterium discharges. These results are consistent with the picture that oxygen gets into the plasma mainly through some process of chemical detachment, but that the metals are produced by neutral-particle sputtering at the walls. This conclusion pertains to discharges that are kept centered in the vacuum chamber so that the plasma limiter interactions are minimized. Under this condition the ion temperature near the edge of the current column is apparently low enough that charged particle sputtering of the limiter is a relatively small effect.
Date: January 1, 1982
Creator: Isler, R.C.; Murray, L.E. & Edmonds, P.H.
Partner: UNT Libraries Government Documents Department

A procedure for generating quantitative 3-D camera views of tokamak divertors

Description: A procedure is described for precision modeling of the views for imaging diagnostics monitoring tokamak internal components, particularly high heat flux divertor components. These models are required to enable predictions of resolution and viewing angle for the available viewing locations. Because of the oblique views expected for slot divertors, fully 3-D perspective imaging is required. A suite of matched 3-D CAD, graphics and animation applications are used to provide a fast and flexible technique for reproducing these views. An analytic calculation of the resolution and viewing incidence angle is developed to validate the results of the modeling procedures. The calculation is applicable to any viewed surface describable with a coordinate array. The Tokamak Physics Experiment (TPX) diagnostics for infrared viewing are used as an example to demonstrate the implementation of the tools. For the TPX experiment the available locations are severely constrained by access limitations at the end resulting images are marginal in both resolution and viewing incidence angle. Full coverage of the divertor is possible if an array of cameras is installed at 45 degree toroidal intervals. Two poloidal locations are required in order to view both the upper and lower divertors. The procedures described here provide a complete design tool for in-vessel viewing, both for camera location and for identification of viewed surfaces. Additionally these same tools can be used for the interpretation of the actual images obtained by the actual diagnostic.
Date: May 1, 1996
Creator: Edmonds, P.H. & Medley, S.S.
Partner: UNT Libraries Government Documents Department

Pumped limiter development on ISX

Description: Pumped limiter configurations are being suggested for FED and INTOR for helium ash exhaust and fuel particle control. The goal of the pump limiter studies in ISX is the selection of the most promising concept and its evaluation in the ISX-C device under the following conditions: (1) quasi steady state operation (less than or equal to 30s), (2) high edge power densities, and (3) particle control by means of mechanical devices. We are considering various options, including particle scraper and ballistic particle collection concepts as well as the current FED design. In ISX-B we will test a full-size pump limiter and directly compare the heat removal and particle control capabilities with a bundle divertor. In ISX-C the steady state operation characteristics of pump limiters will be explored.
Date: January 1, 1981
Creator: Mioduszewski, P.K.; Edmonds, P.H. & Sheffield, J.
Partner: UNT Libraries Government Documents Department

ATF neutral beam injection system

Description: The Advanced Toroidal Facility is a stellarator torsatron being built at Oak Ridge National Laboratory to investigate improved plasma confinement schemes. Plasmas heating will be carried out predominantly by means of neutral beam injection. This paper describes the basic parameters of the injection system. Numerical calculations were done to optimize the aiming of the injectors. The results of these calculations and their implications on the neutral power to the machine are elaborated. The effects of improving the beam optics and altering the focal length on the power transmitted to the plasma are discussed.
Date: January 1, 1985
Creator: Menon, M.M.; Morris, R.N. & Edmonds, P.H.
Partner: UNT Libraries Government Documents Department

Plasma-wall impurity experiments in ISX-A

Description: The ISX-A was a tokamak designed for studying plasma-wall interactions and plasma impurities. It fulfilled this role quite well, producing reliable and reproducible plasmas which had currents up to 175 kA and energy containment times up to 30 msec. With discharge precleaning, Z/sub eff/ was as low as 1.6; with titanium evaporation, Z/sub eff/ approached 1.0. Values of Z/sub eff/ greater than or equal to 2.0 were found to be proportional to residual impurity gases in the vacuum system immediately following a discharge. However, there was no clear dependence of Z/sub eff/ on base pressure. Stainless steel limiters were used in most of the ISX-A experiments. When carbon limiters were introduced into the vacuum system, Z/sub eff/ increased to 5.6. After twelve days of cleanup with tokamak discharges, during which time Z/sub eff/ steadily decreased, the carbon limiters tended to give slightly higher values of Z/sub eff/ than stainless steel limiters. Injection of less than 10/sup 16/ atoms of tungsten into discharges caused the power incident on the wall to double and the electron temperature profile to become hollow.
Date: August 1, 1978
Creator: Colchin, R.J.; Bush, C.E. & Edmonds, P.H.
Partner: UNT Libraries Government Documents Department

Gettering in ISX-B

Description: Gettering is used in the ISX-B tokamak to reduce the impurity concentration. This paper documents the gettering process used, and compares the expected changes in recycling and radiation with those observed experimentally. The enlargement of the operating regime (1/q, anti n/sub e/ R/B/sub phi/ space) is discussed. Finally, the effect on one of the objectives of the experimental program, that of obtaining high values of beta, is described.
Date: January 1, 1982
Creator: Wootton, A.J.; Edmonds, P.H.; Isler, R.C. & Mioduszewski, P.
Partner: UNT Libraries Government Documents Department

Scrapeoff layer studies with an instrumented limiter

Description: An instrumented limiter was used as a diagnostic to measure energy deposition and power fluxes in the scrapeoff layer. The limiter consisted of an array of 12 tiles, each equipped with a thermocouple. This arrangement represented a calorimeter array that yielded the total limiter energy and decay length of the power flux in the scrapeoff layer. Energy deposition and decay lengths are discussed for various plasma parameters.
Date: January 1, 1986
Creator: Mioduszewski, P.; Edmonds, P.H.; Emerson, L.C. & Simpkins, J.E.
Partner: UNT Libraries Government Documents Department

Fabrication of high rate chromium getter sources for fusion applications

Description: Design and fabrication techniques are described for the manufacture of large-capacity chromium getter sources, analogous to the commercially available titanium getter source known as Ti-Ball, manufactured by Varian Associates.
Date: January 1, 1983
Creator: Gabbard, W.A.; Simpkins, J.E.; Mioduszewski, P. & Edmonds, P.H.
Partner: UNT Libraries Government Documents Department

Compact tokamak reactors. Part 1 (analytic results)

Description: We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model.
Date: September 13, 1996
Creator: Wootton, A. J.; Wiley, J. C.; Edmonds, P. H. & Ross, D. W.
Partner: UNT Libraries Government Documents Department

TPX diagnostics for tokamak operation, plasma control and machine protection

Description: The diagnostics for TPX are at an early design phase, with emphasis on the diagnostic access interface with the major tokamak components. Account has to be taken of the very severe environment for diagnostic components located inside the vacuum vessel. The placement of subcontracts for the design and fabrication of the diagnostic systems is in process.
Date: August 1, 1995
Creator: Edmonds, P.H.; Medley, S.S. & Young, K.M.
Partner: UNT Libraries Government Documents Department

Weight change measurements of erosion/deposition at beryllium limiter tiles in ISX-B

Description: The weight changes of Be tiles which functioned as a rail limiter in ISX-B for more than 3500 beam-heated discharges have been determined. The net weight loss for the limiter was 2.0 g, with the central tiles losing a total of 3.2 g and inboard tiles gaining 1.2 g. The weight loss is attributed primarily to the release of Be droplets as a result of limiter surface melting. The weight gains resulted from an inward flow of molten material along the limiter surface. The results indicate high erosion (melt loss) with incomplete and nonuniform redeposition (melt flow) of limiter material during periods of limiter melting.
Date: July 1, 1985
Creator: Roberto, J.B.; Edmonds, P.H.; England, A.C.; Gabbard, A. & Zuhr, R.A.
Partner: UNT Libraries Government Documents Department

Studies on impurity control and hydrogen pumping with chromium gettering in ISX-B

Description: Chromium gettering has been proven to be a trouble-free and efficient method of surface pumping in tokamaks. The impurity control capabilities are excellent and comparable to that of titanium. The hydrogen uptake is reduced to monolayer quantities on the surface. The expansion of the operating space is similar to that seen with titanium without the disadvantage of strongly increased hydrogen fluxes. Possible applications of chromium gettering are: impurity control in contemporary tokamaks; surface pumping in short pulse DT-burning devices to minimize tritium inventory, and wall conditioning of future large machines prior to operation.
Date: January 1, 1984
Creator: Mioduszewski, P.; Simpkins, J.E.; Edmonds, P.H.; Isler, R.C.; Lazarus, E.A.; Ma, C.H. et al.
Partner: UNT Libraries Government Documents Department

Particle removal with pump limiters in ISX-B

Description: First pump limiter experiments were performed on ISX-B. Two pump limiter modules were installed in the top and bottom of one toroidal sector of the tokamak. The modules consist of inertia cooled, TiC coated graphite heads and Zr-Al getter pumps each with a pumping speed of 1000 to 2000 l/s. The objective of the initial experiments was the demonstration of plasma particle control with pump limiters. The first set of experiments were performed in ohmic discharges (OH) in which the effect of the pump limiters on the plasma density was clearly demonstrated. In discharges characterized by: I/sub p/ = 110 kA, B/sub T/ = 15 kG, anti n/sub e/ = 1 - 5 x 10/sup 13/ cm/sup -3/ and t = 0.3 s the pressure rise in the pump limiters was typically 2 mTorr with the pumps off and 0.7 mTorr after activating the pumps. When the pumps were activated, the line-average plasma density decreased by up to a factor 2 at identical gas flow rates. The second set of measurements were performed in neutral beam heated discharges (NBI) with injected powers between 0.6 MW and 1.0 MW. Due to a cooling problem on one of the Zr-Al pumps the NBI experiments were carried out with one limiter only. The maximum pressure observed in NBI-discharges was 5 mTorr without activating the pumps, i.e., approximately twice as high as in OH-discharges. The exhaust efficiency, which is defined as the removed particle flux over the total particle flux in the scrape-off layer is estimated to be 5%.
Date: January 1, 1983
Creator: Mioduszewski, P.; Emerson, L.C.; Simpkins, J.E.; Wootton, A.J.; Bush, C.E.; Carnevali, A. et al.
Partner: UNT Libraries Government Documents Department

High-beta studies in the ISX-B tokamak

Description: Experimental results from the ISX-B tokamak (major radius R/sub 0/=0.93 m, minor radius a=0.26 m, plasma current I/sub p/ 230 kA, elongation k=1.1 to 1.6, toroidal field B/sub phi/ less than or equal to 1.5 T, density anti n/sub e/ less than or equal to 1.1x10/sup 20/ m/sup -3/, neutral-beam power P/sub b/ less than or equal to 2.5 MW) at volume-averaged beta (<..beta..>) values up to 2.5% are described. Two aspects of these studies are presented: (1) empirical scaling of beta and of confinement time, and (2) MHD equilibrium analysis of ISX-B plasmas. The main points which are made are, respectively: (1) global confinement time tau/sub E/ exhibits a strong positive dependence on plasma current (I/sub p//sup 3/2/), a negative dependence on beam power (P/sub b//sup -2/3/), little or no dependence on <..beta..> or density, and no correlation with variations in the observed (m=1;n=1 dominated) MHD activity, and (2) profile analysis is coupled with an MHD equilibrium solver to obtain a model of the plasma consistent with profile, magnetic probe, and soft x-ray data, and with the boundary conditions imposed by the poloidal coil currents.
Date: January 1, 1982
Creator: Neilson, G.H.; Bates, S.C.; Bell, J.D.; Bush, C.E.; Carreras, B.A.; Charlton, L.A. et al.
Partner: UNT Libraries Government Documents Department