WorldWideScience

Sample records for tftr tokamak

  1. Disruptions in the TFTR tokamak

    International Nuclear Information System (INIS)

    Janos, A.; Fredrickson, E.D.; McGuire, K.; Batha, S.H.; Bell, M.G.; Bitter, M.; Budny, R.; Bush, C.E.; Efthimion, P.C.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Jobes, F.C.; Johnson, D.W.; Levinton, F.; Mansfield, D.; Meade, D.; Medley, S.S.; Monticello, D.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Park, H.; Park, W.; Post, D.E.; Schivell, J.; Strachan, J.D.; Taylor, G.; Ulrickson, M.; Goeler, S. von; Wilfrid, E.; Wong, K.L.; Yamada, M.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J.; Drake, J.F.; Kleva, R.G.; Fleischmann, H.H.

    1993-03-01

    For a successful reactor, it will be useful to predict the occurrence of disruptions and to understand disruption effects including how a plasma disrupts onto the wall and how reproducibly it does so. Studies of disruptions on TFTR at both high-β pol and high-density have shown that, in both types, a fast growing m/n=1/1 mode plays an important role. In highdensity disruptions, a newly observed fast m/n = 1/1 mode occurs early in the thermal decay phase. For the first time in TFTR q-profile measurements just prior to disruptions have been made. Experimental studies of heat deposition patterns on the first wall of TFTR due to disruptions have provided information on MHD phenomena prior to or during the disruption, how the energy is released to the wall, and the reproducibility of the heat loads from disruptions. This information is important in the design of future devices such as ITER. Several new processes of runaway electron generation are theoretically suggested and their application to TFTR and ITER is considered, together with a preliminary assessment of x-ray data from runaways generated during disruptions

  2. Three novel tokamak plasma regimes in TFTR

    International Nuclear Information System (INIS)

    Furth, H.P.

    1985-10-01

    Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

  3. Engineering design of TFTR and it's impact on future tokamaks

    International Nuclear Information System (INIS)

    Sabado, M.M.

    1981-01-01

    TFTR is a second generation tokamak whose key objective is scientific break-even. TFTR is expected to be the first machine to demonstrate proper combination of plasma confinement time, density, and temperature to obtain this objective. A summary of major TFTR design parameters, including TFM, is presented, and their potential impact on future tokamaks discussed. Details of the updated engineering design and analysis of components are described. Status of major hardware fabrication, assembly installation and test are reviewed. TFTR features, technology, predicted performance and their potential implication for future tokamaks are summarized

  4. Experimental results from the TFTR tokamak

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Arunasalam, V.; Bell, J.D.

    1986-10-01

    Recent experiments on TFTR have extended the operating regime of TFTR in both ohmic- and neutral-beam-heated discharges. The TFTR tokamak has reached its original machine design specifications (I/sub p/ = 2.5 MA and B/sub T/ = 5.2 T). Initial neutral-beam-heating experiments used up to 6.3 MW of deuterium beams. With the recent installation of two additional beamlines, the power has been increased up to 11 MW. A deuterium pellet injector was used to increase the central density to 2.5 x 10 20 m -3 in high current discharges. At the opposite extreme, by operating at low plasma current (I/sub p/ ∼ 0.8 MA) and low density (anti n/sub e/ ∼ 1 x 10 19 m -3 ), high ion temperatures (9 +- 2 keV) and rotation speeds (7 x 10 5 m/s) have been achieved during injection. In addition, plasma compression experiments have demonstrated acceleration of beam ions from 82 keV to 150 keV, in accord with expectations. The wide operating range of TFTR, together with an extensive set of diagnostics and a flexible control system, has facilitated transport and scaling studies of both ohmic- and neutral-beam-heated discharges. The results of these confinement studies are presented

  5. Design of the TFTR [Tokamak Fusion Test Reactor] maintenance manipulator

    International Nuclear Information System (INIS)

    Loesser, G. D.; Heitzenroeder, P.; Bohme, G.; Selig, M.

    1987-01-01

    The Tokamak Fusion Test Reactor (TFTR) plans to generate a total of 3 x 10 21 neutrons during its deuterium-tritium run period in 1900. This will result in high levels of radiation, especially within the TFTR vacuum vessel. The maintenance manipulator's mission is to assist TFTR in meeting Princeton Plasma Physics Laboratory's personnel radiation exposure criteria and in maintaining as-low-as-reasonably-achievable principals by limiting the radiation exposure received by operating and maintenance personnel. The manipulator, which is currently being fabricated and tested by Kernforschungszentrum Karlsruhe, is designed to perform limited, but routine and necessary, functions within the TFTR vacuum torus after activation levels within the torus preclude such functions being performed by personnel. These functions include visual inspection, tile replacement, housekeeping tasks, diagnostic calibrations, and leak detection. To meet its functional objectives, the TFTR maintenance manipulator is required to be operable in TFTR's very high vacuum environment (typically 2 x 10 -8 Torr). It must also be bakeable at 150 degree C and able to withstand the radiation environment

  6. β limit disruptions in the TFTR tokamak

    International Nuclear Information System (INIS)

    Fredrickson, E.D.; McGuire, K.; Janos, A.; Bell, M.; Budny, R.V.; Bush, C.E.; Manickam, J.; Mynick, H.; Nazikian, R.; Taylor, G.

    1994-11-01

    A disruptive β limit (β = plasma pressure/magnetic pressure) is observed in high performance plasmas in TFTR. The MHD character of these disruptions differs substantially from the disruptions in high density plasmas (density limit disruptions) on TFTR. The high β disruptions can occur with less than a milliseconds warning in the form of a fast growing precursor. The precursor appears to be an external kink or internal (m,n)=(1,1) kink strongly coupled through finite β effects and toroidal terms to higher m components. It does not have the open-quote cold bubble close-quote structure found in density limit disruptions. There is also no evidence for a change in the internal inductance, i.e., a major reconnection of the flux, at the time of the thermal quench

  7. Ohmic Heating System for the TFTR Tokamak

    International Nuclear Information System (INIS)

    Petree, F.; Cassel, R.

    1977-01-01

    The TFTR Ohmic Heating (OH) System will apply 140,000 volt impulses upon the OH coils to start the plasma. In order to reduce the voltage stress to ground on the OH coils to 12 kV without changing the magnetic field induced by the OH system in the plasma, six d-c current interrupters will be applied to six entry points in the OH coil system. And in order to impart a nearly rectangular shape to these impulses, the voltage determining elements will be nonlinear resistances placed in parallel with the interrupters. These nonlinear resistors, made of semiconducting material, are not normally used in repetitive or continuous duty, and their proper functioning is crucial to the reliable operation of the system. The system described herein, is being revised owing to the impact of revisions to the Toroidal Field Coil System, and to refinements to the OH System design

  8. High-Q plasmas in the TFTR tokamak

    International Nuclear Information System (INIS)

    Jassby, D.L.; Barnes, C.W.; Bell, M.G.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.V.; Bush, C.E.; Dylla, H.F.; Efthimion, P.C.; Fredrickson, E.D.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Hsuan, H.; Janos, A.C.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kieras-Phillips, C.; Kilpatrick, S.J.; LaMarche, P.H.; LeBlanc, B.; Mansfield, D.K.; Marmar, E.S.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.K.; Paul, S.F.; Pitcher, S.; Ramsey, A.T.; Redi, M.H.; Sabbagh, S.A.; Scott, S.D.; Snipes, J.; Stevens, J.; Strachan, J.D.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Terry, J.L.; Timberlake, J.R.; Towner, H.H.; Ulrickson, M.; von Goeler, S.; Wieland, R.M.; Williams, M.; Wilson, J.R.; Wong, K.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J.

    1991-01-01

    In the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], the highest neutron source strength S n and D--D fusion power gain Q DD are realized in the neutral-beam-fueled and heated ''supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, S n increases approximately as P 1.8 b . The highest-Q shots are characterized by high T e (up to 12 keV), T i (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad T e profiles, and lower Z eff . Replacement of critical areas of the graphite limiter tiles with carbon-fiber composite tiles and improved alignment with the plasma have mitigated the ''carbon bloom.'' Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, Q DD increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness [n e (0)/left-angle n e right-angle] during the beam pulse. To date, the best fusion results are S n =5x10 16 n/sec, Q DD =1.85x10 -3 , and neutron yield=4.0x10 16 n/pulse, obtained at I p =1.6--1.9 MA and beam energy E b =95--103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50%--60% of S n arises from beam--target reactions, with the remainder divided between beam--beam and thermonuclear reactions, the thermonuclear fraction increasing with P b

  9. Neutron spectroscopy on TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Nishitani, T.; Strachan, J.D.

    1988-05-01

    This paper describes the use of an 3 He ionization chamber for neutron spectroscopy on TFTR during 1987. The ion temperature was measured using neutron spectroscopy for one set of ohmically heated plasmas. The deduced ion temperatures agreed to within 20% with those measured by other diagnostics. 11 refs., 11 figs., 1 tab

  10. Phenomenology of high density disruptions in the TFTR tokamak

    International Nuclear Information System (INIS)

    Fredrickson, E.D.; McGuire, K.M.; Bell, M.G.

    1993-01-01

    Studies of high density disruptions on TFTR, including a comparison of minor and major disruptions at high density, provide important new information regarding the nature of the disruption mechanism. Further, for the first time, an (m,n)=(1,1) 'cold bubble' precursor to high density disruptions has been experimentally observed in the electron temperature profile. The precursor to major disruptions resembles the 'vacuum bubble' model of disruptions first proposed by B.B. Kadomtsev and O.P. Pogutse (Sov. Phys. - JETP 38 (1974) 283). (author). Letter-to-the-editor. 25 refs, 3 figs

  11. TFTR/JET INTOR workshop on plasma transport tokamaks

    International Nuclear Information System (INIS)

    Singer, C.E.

    1985-01-01

    This report summarizes the proceedings of a Workshop on transport models for prediction and analysis of tokamak plasma confinement. Summaries of papers on theory, predictive modeling, and data analysis are included

  12. Fusion reactivity, confinement, and stability of neutral-beam heated plasmas in TFTR and other tokamaks

    International Nuclear Information System (INIS)

    Park, Hyeon, K.

    1996-05-01

    The hypothesis that the heating beam fueling profile shape connects the edge condition and improved core confinement and fusion reactivity is extensively studied on TFTR and applied to other tokamaks. The derived absolute scalings based on beam fueling profile shape for the stored energy and neutron yield can be applied to the deuterium discharges at different major radii in TFTR. These include Supershot, High poloidal beta, L-mode, and discharges with a reversed shear (RS) magnetic configuration. These scalings are also applied to deuterium-tritium discharges. The role of plasma parameters, such as plasma current, Isdo2(p), edge safety factor, qsdo5(a), and toroidal field, Bsdo2(T), in the performance and stability of the discharges is explicitly studied. Based on practical and externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path for a discharge condition with fusion power gain, Q > 1 is suggested based on this study. Similar physics interpretation is provided for beam heated discharges on other major tokamaks

  13. Vacuum system for the tokamak fusion test reactor (TFTR)

    International Nuclear Information System (INIS)

    Lange, W.J.; Green, D.; Sink, D.A.

    1976-01-01

    The vacuum system for TFTR is described. Insofar as possible, conventional and ultrahigh vacuum (UHV) components and technology will be employed. Subassemblies will be prebaked in vacuum to reduce subsequent outgassing, and assembly will employ TIG welding and metal gaskets. It is not anticipated that the totally assembled torus with its numerous diagnostic appendages will be baked in situ to a high temperature, however a lower bakeout temperature (approximately 250 0 C) is under consideration. Final vacuum conditioning will be performed using discharge cleaning to obtain a specific outgassing rate of less than or = to 10 -10 Torr liter/sec cm 2 hydrogen isotopes and less than or = to 10 -12 Torr liter/sec cm 2 of other gases, and a base pressure of less than or = to 5 x 10 -8 Torr

  14. A Michelson interferometer/polarimeter on the Tokamak Fusion Test Reactor (TFTR)

    International Nuclear Information System (INIS)

    Park, H.K.; Mansfield, D.K.; Johnson, L.C.; Ma, C.H.

    1987-01-01

    A multichannel interferometer/polarimeter for the Tokamak Fusion Test Reactor (TFTR) has been developed in order to study the time dependent plasma current density (J/sub p/) and electron density (n/sub e/) profile simultaneously. The goal of the TFTR is demonstration of breakeven via dueuterium and tritium (DT) plasma. In order to be operated and maintained during DT operation phase, the system is designed based on the Michelson geometry which possesses intrinsic standing wave problems. So far, there has been no observable signals due to these standing waves. However, a standing wave resulted from the beam path design to achieve a optimum use of the laser power was found. This standing wave has not prevented initial 10 channel interferometer operation. However, a single channel polarimeter test indicated this standing wave was fatal for Faraday notation measurements. Techniques employing 1/2 wave plates and polarizers have been applied to eliminate this standing wave problem. The completion of 10 channel Faraday rotation measurements may be feasible in the near future

  15. Electron temperature profiles in high power neutral-beam-heated TFTR [Tokamak Fusion Test Reactor] plasmas

    International Nuclear Information System (INIS)

    Taylor, G.; Grek, B.; Stauffer, F.J.; Goldston, R.J.; Fredrickson, E.D.; Wieland, R.M.; Zarnstorff, M.C.

    1987-09-01

    In 1986, the maximum neutral beam injection (NBI) power in the Tokamak Fusion Test Reactor (TFTR) was increased to 20 MW, with three beams co-parallel and one counter-parallel to I/sub p/. TFTR was operated over a wide range of plasma parameters; 2.5 19 19 m -3 . Data bases have been constructed with over 600 measured electron temperature profiles from multipoint TV Thomson scattering which span much of this parameter space. We have also examined electron temperature profile shapes from electron cyclotron emission at the fundamental ordinary mode and second harmonic extraordinary mode for a subset of these discharges. In the light of recent work on ''profile consistency'' we have analyzed these temperature profiles in the range 0.3 < (r/a) < 0.9 to determine if a profile shape exists which is insensitive to q/sub cyl/ and beam-heating profile. Data from both sides of the temperature profile [T/sub e/(R)] were mapped to magnetic flux surfaces [T/sub e/(r/a)]. Although T/sub e/(r/a), in the region where 0.3 < r/a < 0.9 was found to be slightly broader at lower q/sub cyl/, it was found to be remarkably insensitive to β/sub p/, to the fraction of NBI power injected co-parallel to I/sub p/, and to the heating profile going from peaked on axis, to hollow. 10 refs., 8 figs

  16. Long- and short-term trends in vessel conditioning of TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    LaMarche, P.H.; Dylla, H.F.; Bell, M.G.

    1986-10-01

    We have investigated trends in the conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel during the May 1984 to April 1985 run period. The initial conditioning of the vessel, consisting of glow discharge cleaning (GDC) and pulse discharge cleaning (PDC) in concert with a 150 0 C vessel bakeout, is necessary to assure plasma operation after atmospheric venting. A long-term conditioning process, ascribed to limiter conditioning, effectively improves operational conditions during the course of the run. Over several thousand high power plasma discharges, the improvement was documented by using standard parameter (fiducial) plasma discharges. Several techniques demonstrated short-term improvements in vessel conditioning during this time period, including: Cr gettering and programming the plasma position relative to the limiter contact area

  17. Dielectronic satellite spectra of hydrogenlike iron from TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Decaux, V.; Bitter, M.; Hsuan, H.; von Goeler, S.; Hill, K.W.; Hulse, R.A.; Taylor, G.; Park, H.; Bhalla, C.P.

    1990-08-01

    Spectra of hydrogenlike iron, Fe26, have been observed from Tokamak Fusion Test Reactor (TFTR) plasmas with a high-resolution crystal spectrometer. The experimental arrangement permits simultaneous observation of the Fe26 Ly-α 1 and Ly-α 2 lines and the associated dielectronic satellites, which are due to transitions 1snl-2pnl' with n ≥ 2, as well as the heliumlike 1s 2 ( 1 S 0 )-1s4p( 1 P 1 )and both hydrogenlike Ly-β 1 and Ly-β 2 lines from chromium. Relative wavelengths and line intensities can be determined very accurately. The spectral data are in very good agreement with theoretical calculations. The observed spectra have also been used to estimate the total dielectronic recombination rate coefficient of Fe26. 30 refs., 4 figs., 3 tabs

  18. MIRI: A multichannel far-infrared laser interferometer for electron density measurements on TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Park, H.K.; Johnson, L.C.; Anderson, H.M.; Chouinard, R.; Foote, V.S.; Ma, C.H.; Clifton, B.J.

    1987-07-01

    A ten-channel far-infrared laser interferometer which is routinely used to measure the spatial and temporal behavior of the electron density profile on the TFTR tokamak is described and representative results are presented. This system has been designed for remote operation in the very hostile environment of a fusion reactor. The possible expansion of the system to include polarimetric measurements is briefly outlined. 13 refs., 8 figs

  19. Neutron sawtooth behavior in the PLT, DIII-D, and TFTR tokamaks

    International Nuclear Information System (INIS)

    Lovberg, J.A.; Heidbrink, W.W.; Strachan, J.D.; Zaveryaev, V.S.

    1988-10-01

    The effect of the sawtooth instability on the 2.5 MeV neutron emission in the PLT, DIII-D, and TFTR tokamaks is studied. In thermonuclear plasmas, the instability typically results in a 20% reduction in emission. The time evolution of the thermonuclear neutron signal suggests that the sawtooth crash consists of four phases. First, the electron density profile flattens rapidly (in /approximately/30μsec on PLT) but, in some cases, there is little associated change in neutron emission, suggesting that most reacting ions remain confined in the sawtooth region but do not completely mix. After the electron sawtooth, the ions continue to mix, resulting in a /approximately/10% reduction in neutron emission in /approximately/0.5 msec. The emission then decays more slowly during the final two phases. Thermalization of reacting ions on a /approximately/3/tau//sub ii/ time scale accounts for only /approximately/20% of the slow drop. Most of the slow drop seems to be caused by loss of ion energy from the mixing region (an ion heat pulse). 36 refs., 15 figs., 1 tabs

  20. Helium, iron and electron particle transport and energy transport studies on the TFTR tokamak

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Efthimion, P.C.; Rewoldt, G.; Stratton, B.C.; Tang, W.M.; Grek, B.; Hill, K.W.; Hulse, R.A.; Johnson, D.W.; Mansfield, D.K.; McCune, D.; Mikkelsen, D.R.; Park, H.K.; Ramsey, A.T.; Redi, M.H.; Scott, S.D.; Taylor, G.; Timberlake, J.; Zarnstorff, M.C.

    1993-03-01

    Results from helium, iron, and electron transport on TFTR in L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the Supershot The helium profile shape is consistent with predictions from quasilinear electrostatic drift-wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the Supershot, differences are found in the L-Mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of three larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor

  1. Helium, Iron and Electron Particle Transport and Energy Transport Studies on the TFTR Tokamak

    Science.gov (United States)

    Synakowski, E. J.; Efthimion, P. C.; Rewoldt, G.; Stratton, B. C.; Tang, W. M.; Grek, B.; Hill, K. W.; Hulse, R. A.; Johnson, D .W.; Mansfield, D. K.; McCune, D.; Mikkelsen, D. R.; Park, H. K.; Ramsey, A. T.; Redi, M. H.; Scott, S. D.; Taylor, G.; Timberlake, J.; Zarnstorff, M. C. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Kissick, M. W. (Wisconsin Univ., Madison, WI (United States))

    1993-03-01

    Results from helium, iron, and electron transport on TFTR in L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the Supershot The helium profile shape is consistent with predictions from quasilinear electrostatic drift-wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the Supershot, differences are found in the L-Mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of three larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor.

  2. End points in discharge cleaning on TFTR (Tokamak Fusion Test Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, D.; Dylla, H.F.; Bell, M.G.; Blanchard, W.R.; Bush, C.E.; Gettelfinger, G.; Hawryluk, R.J.; Hill, K.W.; Janos, A.C.; Jobes, F.C.

    1989-07-01

    It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure and aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a) /approx/ 3 at /approx/ 250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H/sub 2/O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of x-rays from runaway electrons, is an indication that TDC is no longer necessary after /approx/ 10/sup 5/ pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150/degree/C to about 250/degree/C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H/sub 2/O is removed from the graphite limiter. 14 refs., 3 figs., 1 tab.

  3. End points in discharge cleaning on TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Mueller, D.; Dylla, H.F.; Bell, M.G.

    1989-07-01

    It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure and aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a) approx 3 at approx 250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H 2 O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of x-rays from runaway electrons, is an indication that TDC is no longer necessary after approx 10 5 pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150 degree C to about 250 degree C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H 2 O is removed from the graphite limiter. 14 refs., 3 figs., 1 tab

  4. Stack and area tritium monitoring systems for the tokamak fusion test reactor (TFTR)

    International Nuclear Information System (INIS)

    Pearson, G.G.; Meixler, L.D.; Sirsingh, R.A.P.

    1992-01-01

    This paper reports on the TFTR Tritium Stack and Area Monitoring Systems which have been developed to provide the required level of reliability in a cost effective manner consistent with the mission of the Tritium Handling System on TFTR. Personnel protection, environmental responsibility, and tritium containing system integrity have been the considerations in system design. During the Deuterium-Tritium (D-T) experiments on TFTR, tritium will be used for the first time as one of the fuels. Area monitors provide surveillance of the air in various rooms at TFTR. Stack monitors monitor the air at the TFTR test site that is exhausted through the HVAC systems, from the room exhaust stacks and the tritium systems process vents. The philosophies for the implementation of the Stack and Area Tritium Monitoring Systems at TFTR are to use hardwired controls wherever personnel protection is involved, and to take advantage of modern intelligent controllers to provide a distributed system to support the functions of tracking, displaying, and archiving concentration levels of tritium for all of the monitored areas and stacks

  5. Characteristics of radiated power for various TFTR [Tokamak Fusion Test Reactor] regimes

    International Nuclear Information System (INIS)

    Bush, C.E.; Schivell, J.; McNeill, D.H.

    1988-04-01

    Power loss studies were carried out to determine the impurity radiation and energy transport characteristics of various TFTR operation and confinement regimes including L-Mode, detached plasma, co-only neutral beam injection (energetic ion regime), and the enhanced confinement (''supershot'') regime. Combined bolometric, spectroscopic, and infrared photometry measurements provide a picture of impurity behavior and power accounting in TFTR. The purpose of this paper is to make a survey of the various regimes with the aim of determining the radiated power signatures of each. 10 refs., 6 figs., 1 tab

  6. In-situ calibration of TFTR [Tokamak Fusion Test Reactor] neutron detectors

    International Nuclear Information System (INIS)

    Hendel, H.W.; Palladino, R.W.; Barnes, C.W.; Diesso, M.; Felt, J.S.; Jassby, D.L.; Johnson, L.C.; Ku, L.P.; Liu, Q.P.; Motley, R.W.; Murphy, H.B.; Murphy, J.; Nieschmidt, E.B.; Roberts, J.A.; Saito, T.; Strachan, J.D.; Waszazak, R.J.; Young, K.

    1990-03-01

    We report results of the TFTR fission detector calibration performed in December 1988. A NBS-traceable, remotely controlled 252 Cf neutron source was moved toroidally through the TFTR vacuum vessel. Detection efficiencies for two 235 U detectors were measured for 930 locations of the neutron point source in toroidal scans at 16 different major radii and vertical heights. These scans effectively simulated the volume-distributed plasma neutron source, and the volume-integrated detection efficiency was found to be insensitive to plasma position. The Campbell mode is useful due to its large overlap with the count rate mode and large dynamic range. The resulting absolute plasma neutron source calibration has an uncertainty of ± 13%. 21 refs., 23 figs., 4 tabs

  7. Assessment of eddy current effects on compression experiments in the TFTR tokamak

    International Nuclear Information System (INIS)

    Wong, K.L.; Park, W.

    1986-05-01

    The eddy current induced on the TFTR vacuum vessel during compression experiments is estimated based on a cylindrical model. It produces an error magnetic field that generates magnetic islands at the rational magnetic surfaces. The widths of these islands are calculated and found to have some effect on electron energy confinement. However, resistive MHD simulation results indicate that the island formation process can be slowed down by plasma rotation

  8. Mechanical engineering aspects of TFTR

    International Nuclear Information System (INIS)

    Citrolo, J.C.

    1983-04-01

    This paper briefly presents the principles which characterize a tokamak and discusses the mechanical aspects of TFTR, particularly the toroidal field coils and the vacuum chamber, in the context of being key components common to all tokamaks. The mechanical loads on these items as well as other design requirements are considered and the solutions to these requirements as executed in TFTR are presented. Future technological developments beyond the scope of TFTR, which are necessary to bring the tokamak concept to a full fusion-power system, are also presented. Additional methods of plasma heating, current drive, and first wall designs are examples of items in this category

  9. TFTR DT preparation project status

    Energy Technology Data Exchange (ETDEWEB)

    Perry, E.D.; Dudek, L.E.

    1993-11-01

    The Tokamak Fusion Test Reactor (TFTR) research program is preparing to commence the first high power Deuterium-Tritium (DT) experiments of the US Fusion Program. Hardware upgrades to TFTR required for DT operations have been completed. This paper discusses these hardware preparations.

  10. Handling of tritium at TFTR

    International Nuclear Information System (INIS)

    Pierce, C.W.; Howe, H.J.; Yemin, L.; Lind, K.

    1977-01-01

    Some of the engineering approaches taken at TFTR for the tritium control systems are discussed as the requirements being placed on the tritium systems by the operating scenarios of the Tokamak. The tritium control systems presently being designed for TFTR will limit the annual release to the environment to less than 100 curies

  11. Gas utilization in TFTR [Tokamak Fusion Test Reactor] neutral beam injectors

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Gammel, G.M.; Kugel, H.W.; Grisham, L.R.; Stevenson, T.N.; von Halle, A.; Williams, M.D.

    1987-08-01

    Measurements of gas utilization in a test TFTR neutral beam injector have been performed to study the feasibility of running tritium neutral beams with existing ion sources. Gas consumption is limited by the restriction of 50,000 curies of T 2 allowed on site. It was found that the gas efficiency of the present long-pulse ion sources is higher than it was with previous short-pulse sources. Gas efficiencies were studied over the range of 35 to 55%. At the high end of this range the neutral fraction of the beam fell below that predicted by room temperature molecular gas flow. This is consistent with observations made on the JET injectors, where it has been attributed to beam heating of the neutralizer gas and a concomitant increase in conductance. It was found that a working gas isotope exchange from H 2 to D 2 could be accomplished on the first beam shot after changing the gas supply, without any intermediate preconditioning. The mechanism believed responsible for this phenomenon is heating of the plasma generator walls by the arc and a resulting thermal desorption of all previously adsorbed and implanted gas. Finally, it was observed that an ion source conditioned to 120 kV operation could produce a beam pulse after a waiting period of fourteen hours by preceding the beam extraction with several hi-pot/filament warm-up pulses, without any gas consumption. 18 refs., 7 figs., 2 tabs

  12. Ion temperature gradient driven transport in a density modification experiment on the TFTR tokamak

    International Nuclear Information System (INIS)

    Horton, W.; Lindberg, D.; Kim, J.Y.; Dong, J.Q.; Hammett, G.W.; Scott, S.D.; Zarnstorff, M.C.; Hamaguchi, S.

    1991-07-01

    TFTR profiles from a supershot density-modification experiment are analyzed for their local and ballooning stability to toroidal η i -modes in order to understand the initially puzzling results showing no increase in X i when a pellet is used to produce an abrupt and large increase in the η i parameter. The local stability analysis assumes that k parallel = 1/qR and ignores the effects of shear, but makes no assumption on the magnitude of k parallel v ti /ω. The ballooning stability analysis determines a self-consistent linear spectrum of k parallel's including the effect of shear and toroidicity, but it expands in k parallel v ti /ω ≤ 1, which is a marginal assumption for this experiment. Nevertheless, the two approaches agree well and show that the mixing length estimate of the transport rate does not change appreciably during the density-modification and has a value close to or less than the observed X i , in contrast to most previous theories which predicted X i 's which were over an order-of-magnitude too large. However, we are still unable to explain the observed increase X i (r) with minor radius by adding the effects of the finite beta drift - MHD mode coupling, the slab-like mode, or the trapped electron response. The experimental tracking 0.2 e /X i i and trapped-electron driving mechanisms are operating. 4 refs., 5 figs., 5 tabs

  13. Cryosorption of helium on argon frost TFTR [Tokamak Fusion Test Reactor] neutral beamlines

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Cropper, M.B.; Dylla, H.F.

    1989-11-01

    Helium pumping on argon frost has been investigated on TFTR neutral beam injectors and shown to be viable for limited helium beam operation. Maximum pumping speeds are ∼ 25% less than those measured for pumping of deuterium. Helium pumping efficiency is low, > 20 argon atoms are required to pump each helium atom. Adsorption isotherms are exponential and exhibit a two-fold increase in adsorption capacity as the cryopanel temperature is reduced from 4.3 K to 3.7 K. Pumping speed was found to be independent of cryopanel temperature over the temperature range studied. After pumping a total of 2000 torr-l of helium, the beamline base pressure rose to 2x10 -5 torr from an initial value of 10 -8 torr. Accompanying this three order of magnitude increase in pressure was a modest 40% decrease in pumping speed. The introduction of 168 torr-l of deuterium prior to helium injection reduced the pumping speed by a factor of two with no decrease in adsorption capacity. 29 refs., 7 figs

  14. Local transport barrier formation and relaxation in reverse-shear plasmas on the TFTR tokamak

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.A.

    1997-02-01

    The roles of turbulence stabilization by sheared E x B flow and Shafranov-shift gradients are examined for TFTR. Enhanced Reverse-Shear plasmas. Both effects in combination provide the basis of a positive-feedback model that predicts reinforced turbulence suppression with increasing pressure gradient. Local fluctuation behavior at the onset of ERS confinement is consistent with this framework. The power required for transitions into the ERS regime are lower when high power neutral beams are applied earlier in the current profile evolution, consistent with the suggestion that both effects play a role. Separation of the roles of E x B and Shafranov shift effects was performed by varying the E x B shear through changes in the toroidal velocity with nearly-steady-state pressure profiles. Transport and fluctuation levels increase only when E x B shearing rates are driven below a critical value that is comparable to the fastest linear growth rates of the dominant instabilities. While a turbulence suppression criterion that involves the ratio of shearing to linear growth rates is in accord with many of these results, the existence of hidden dependencies of the criterion is suggested in experiments where the toroidal field was varied. The forward transition into the ERS regime has also been examined in strongly rotating plasmas. The power threshold is higher with unidirectional injection than with balanced injection

  15. Status report on TFTR

    International Nuclear Information System (INIS)

    Reardon, P.J.

    1978-01-01

    The primary objectives of the TFTR are the generation and confinement of 5 to 10 keV (50 to 100 million degrees) reactor-grade plasmas in a tokamad magnetic-field configuration, and the production of fusion energy on a pulsed basis, from the reaction of deuterum and tritium. The TFTR will be used to study the physics of burning plasmas and the engineering aspects of a D-T burning tokamak operating with reactor-level plasma conditions. The overall TFTR program is intended to produce scientific and technical information, component hardware, and the design, construction, and operating experience necessary as input for the future design, construction, and operation of ignition and experimental fusion power reactors. In a very real way the TFTR is prototypical of an Experimenta Power Reactor

  16. Temporal behavior of neutral particle fluxes in TFTR [Tokamak Fusion Test Reactor] neutral beam injectors

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Gammel, G.M.; Roquemore, A.L.

    1989-09-01

    Data from an E parallel B charge exchange neutral analyzer (CENA), which views down the axis of a neutral beamline through an aperture in the target chamber calorimeter of the TFTR neutral beam test facility, exhibit two curious effects. First, there is a turn-on transient lasting tens of milliseconds having a magnitude up to three times that of the steady-state level. Second, there is a 720 Hz, up to 20% peak-to-peak fluctuation persisting the entire pulse duration. The turn-on transient occurs as the neutralizer/ion source system reaches a new pressure equilibrium following the effective ion source gas throughput reduction by particle removal as ion beam. Widths of the transient are a function of the gas throughput into the ion source, decreasing as the gas supply rate is reduced. Heating of the neutalizer gas by the beam is assumed responsible, with gas temperature increasing as gas supply rate is decreased. At low gas supply rates, the transient is primarliy due to dynamic changes in the neutralizer line density and/or beam species composition. Light emission from the drift duct corroborate the CENA data. At high gas supply rates, dynamic changes in component divergence and/or spatial profiles of the source plasma are necessary to explain the observations. The 720 Hz fluctuation is attributed to a 3% peak-to-peak ripple of 720 Hz on the arc power supply amplified by the quadratic relationship between beam divergence and beam current. Tight collimation by CENA apertures cause it to accept a very small part of the ion source's velocity space, producing a signal linearly proportional to beam divergence. Estimated fluctuations in the peak power density delivered to the plasma under these conditions are a modest 3--8% peak to peak. The efffects of both phenomena on the injected neutral beam can be ameliorated by careful operion of the ion sources. 21 refs., 11 figs., 2 tabs

  17. Search for diffusion of counter-passing MeV ions in the TFTR tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Boivin, R.; Chang, C.S.; Hammett, G.; Mynick, H.E.

    1991-07-01

    Confinement studies of MeV ions will play an important role in the research leading to burning plasmas in tokamaks, since any significant radial transport of MeV alpha particles will affect the heating rate or heating profiles of these plasmas. Because the energy, gyroradius, and collisionality of these MeV ions is very different from that of the background plasma, their transport rates cannot be assumed equal to those of the bulk plasma ions. Note that the desired confinement time for 3.5 MeV alphas is set by their thermalization time, which can be up to τ th,α ∼1 sec for the steady-state phase of ITER, requiring D 2 /sec. This is equivalent to over ∼100,000 alpha particle transits of the torus. 28 refs., 24 figs., 2 tabs

  18. Synchronization of timing systems on TFTR

    International Nuclear Information System (INIS)

    Montague, J.; Sichta, P.

    1992-01-01

    This paper reports on the TOKAMAK Fusion Test Reactor (TFTR) facility clock system which has four related timing subsystems: the TFTR shot clock, the Neutral Beams clocks, the Ion Cyclotron Range of Frequencies (ICRF) system clock, and the Disruption Trigger System. These systems have been integrated to support increasingly fast sampling rates in data acquisition and greater accuracy in the firing of the Neutral Beams and ICRF systems during TFTR shots

  19. A comparative study of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.; Bell, R.E.

    2001-01-01

    Confinement bifurcations and subsequent plasma dynamics in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Observations suggest a framework in which ExB shear plays a dominant role in the barrier dynamics. In TFTR, bifurcations from the reverse shear (RS) into the enhanced reverse shear (ERS) regime with high power balanced neutral beam heating (above 25 MW at 4.8 T) resemble edge H mode transitions observed on DIII-D. In both, radial electric field (E r ) excursions precede confinement changes and are manifest as localized changes in the impurity poloidal rotation. Reduced transport follows the excursions, and in both cases strong E r shear is reinforced by the plasma pressure. These characteristics are contrasted with DIII-D negative central shear (NCS) barrier evolution with unidirectional beam injection. There, the improved confinement region can develop slowly, depending on the neutral beam input power and torque. Rapid expansion and deepening of this region follows an increase in the neutral beam heating power. The initial formation phase is modulated by confinement steps and interruptions. An analog for these steps is found in TFTR RS plasmas. Although these do not dominate the TFTR plasma evolution during low power (7 MW) heating, they can represent significant transport reductions when additional heating is applied. In both devices, no strong excursion in E r precedes these latter confinement bifurcations. The triggering event of these steps may be related to current profile relaxation, but it is not always connected with simple integral or half-integer values of the minimum in the q profile. Finally, variations of E r and the ExB shear through the application of unidirectional injection on TFTR yielded plasmas with confinement characteristics and barrier dynamics similar to those of DIII-D NCS plasmas. The data underscore that the physics responsible for the enhanced confinement states is fundamentally

  20. A comparative study of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.A.; Bell, R.E.

    1999-01-01

    Confinement bifurcations and subsequent plasma dynamics in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Observations suggest a framework in which ExB shear plays a dominant role in the barrier dynamics. In TFTR, bifurcations from the reverse shear (RS) into the enhanced reverse shear (ERS) regime with high power balanced neutral beam heating (above 25 MW at 4.8 T) resemble edge H mode transitions observed on DIII-D. In both, radial electric field (E r ) excursions precede confinement changes and are manifest as localized changes in the impurity poloidal rotation. Reduced transport follows the excursions, and in both cases strong E r shear is reinforced by the plasma pressure. These characteristics are contrasted with DIII-D negative central shear (NCS) barrier evolution with unidirectional beam injection. There, the improved confinement region can develop slowly, depending on the neutral beam input power and torque. Rapid expansion and deepening of this region follows an increase in the neutral beam heating power. The initial formation phase is modulated by confinement steps and interruptions. An analog for these steps is found in TFTR RS plasmas. Although these do not dominate the TFTR plasma evolution during low power (7 MW) heating, they can represent significant transport reductions when additional heating is applied. In both devices, no strong excursion in E r precedes these latter confinement bifurcations. The triggering event of these steps may be related to current profile relaxation, but it is not always connected with simple integral or half-integer values of the minimum in the q profile. Finally, variations of E r and the ExB shear through the application of unidirectional injection on TFTR yielded plasmas with confinement characteristics and barrier dynamics similar to those of DIII-D NCS plasmas. The data underscore that the physics responsible for the enhanced confinement states is fundamentally

  1. Compilation of TFTR materials data

    International Nuclear Information System (INIS)

    Havener, W.J.

    1975-12-01

    In order to document the key thermophysical property data used in the conceptual design of Tokamak Fusion Test Reactor (TFTR) systems and components, a series of data packages has been prepared. It is expected that data for additional materials will be added and the information already provided will be updated to provide a project-wide data base

  2. Broadband measurements of electron cyclotron emission in TFTR [Tokamak Fusion Test Reactor] using a quasi-optical light collection system and a polarizing Michelson interferometer

    International Nuclear Information System (INIS)

    Stauffer, F.J.; Boyd, D.A.; Cutler, R.C.; Diesso, M.; McCarthy, M.P.; Montague, J.; Rocco, R.

    1988-04-01

    For the past three years, a Fourier transform spectrometer diagnostic system, employing a fast-scanning polarizing Michelson interferometer, has been operating on the TFTR tokamak at Princeton Plasma Physics Laboratory. It is used to measure the electron cyclotron emission spectrum over the range 2.5 to 18 cm/sup /minus/1/ (75-540 GHz) with a resolution of 0.123 cm/sup /minus/1/(3.7 GHz), at a rate of 72 spectra per second. The quasi-optical system for collecting the light and transporting it through the interferometer to the detector has been designed using the concepts of both Gaussian and geometrical optics in order to produce a system that is efficient over the entire spectral range. The commerical Michelson interferometer was custom-made for this project and is at the state of the art for this type of specialized instrument. Various pre-installation and post-installation tests of the optical system and the interferometer were performed and are reported here. An error propagation analysis of the absolute calibration process is given. Examples of electron cyclotron emission spectra measured in two polarization directions are given, and electron temperature profiles derived from each of them are compared. 34 refs., 17 figs

  3. Development of large high-voltage pressure insulators for the Princeton TFTR [Tokamak Fusion Test Reactor] flexible transmission lines

    International Nuclear Information System (INIS)

    Scalise, D.T.; Fong, E.; Haughian, J.; Prechter, R.

    1986-10-01

    Specially formulated insulator materials with improved strength and high-voltage properties were developed and used for critical components of the flexible transmission lines to the TFTR neutral beam ion sources. These critical components are plates which support central conductors as they exit the high-voltage power supply and enter the ion source enclosure. Each plate acts both as a high-voltage insulator and as a pressure barrier to the SF 6 insulating gas. The original plate was made of commercial glass-epoxy laminate which limited the plate voltage capacity. The newly developed insulator is made of specially-formulated cycloalphatic Di-epoxide whose isotropic properties exhibit increased arc resistance. It is cast in one piece with skirts which greatly increase the breakdown voltage. This paper discusses the design, fabrication and testing of the new insulator

  4. The ICRF antennas for TFTR

    International Nuclear Information System (INIS)

    Hoffman, D.J.; Colestock, P.L.; Gardner, W.L.; Hosea, J.C.; Nagy, A.; Stevens, J.; Swain, D.W.; Wilson, J.R.

    1988-01-01

    Two compact loop antennas have been designed to provide ion cyclotron resonant frequency (ICRF) heating for TFTR. The antennas can convey a total of 10 MW to accomplish core heating in either high-density or high-temperature plasmas. The near-term goal of heating TFTR plasmas and the longer-term goals of ease in handling (for remote maintenance) and high reliability (in an inaccessible tritium tokamak environment) were major considerations in the antenna designs. The compact loop configuration facilitates handling because the antennas fit completely through their ports. Conservative design and extensive testing were used to attain the reliability required for TFTR. This paper summarizes how these antennas will accomplish these goals. 5 figs, 1 tab

  5. TFTR tritium operations lessons learned

    International Nuclear Information System (INIS)

    Gentile, C.A.; Raftopoulos, S.; LaMarche, P.

    1996-01-01

    The Tokamak Fusion Test Reactor which is the progenitor for full D-T operating tokamaks has successfully processed > 81 grams of tritium in a safe and efficient fashion. Many of the fundamental operational techniques associated with the safe movement of tritium through the TFTR facility were developed over the course of many years of DOE tritium facilities (LANL, LLNL, SRS, Mound). In the mid 1980's The Tritium Systems Test Assembly (TSTA) at LANL began reporting operational techniques for the safe handling of tritium, and became a major conduit for the transfer of safe tritium handling technology from DOE weapons laboratories to non-weapon facilities. TFTR has built on many of the TSTA operational techniques and has had the opportunity of performing and enhancing these techniques at America's first operational D-T fusion reactor. This paper will discuss negative pressure employing 'elephant trunks' in the control and mitigation of tritium contamination at the TFTR facility, and the interaction between contaminated line operations and Δ pressure control. In addition the strategy employed in managing the movement of tritium through TFTR while maintaining an active tritium inventory of < 50,000 Ci will be discussed. 5 refs

  6. Operations and maintenance plans for the TFTR

    International Nuclear Information System (INIS)

    Allen, H.L.; Fedor, B.J.

    1978-01-01

    Princeton University Plasma Physics Laboratory (PPPL) is constructing a Tokamak Fusion Test Reactor (TFTR) scheduled to begin operation for fusion research experiments in late 1981, first with hydrogen and deuterium plasmas and later, in the second phase, using tritium for high power fusion studies. This latter mode will introduce considerable complexity to operation and maintenance of the TFTR in terms of meeting requirements for tritium handling, adequate radiation shielding, and corrective and preventive maintenance procedures. In this paper we discuss plans for the installation and preoperational testing of the major subsystems of TFTR, proposed start-up and operating scenarios for the device and the system of operational control. In addition, the TFTR Maintenance Plan and related procedures for specific major maintenance tasks are described, including the use of remote handling equipment and remote manipulators. Each of these topics is addressed in terms of the current status of planning and development

  7. TFTR initial operations

    International Nuclear Information System (INIS)

    Young, K.M.; Bell, M.; Blanchard, W.R.

    1984-01-01

    TFTR (Tokamak Fusion Test Reactor) has operated since December 1982 with ohmically heated plasmas. Routine operation with feedback control of plasma current, position and density has been obtained for plasmas with Isub(p) approx.= 800 kA, a = 68 cm, R = 250 cm, and Bsub(t) = 27 kG. A maximum plasma current of 1 MA was achieved with q approx.= 2.5. Energy confinement times of approx. 150 msec were measured for hydrogen and deuterium plasmas with n-barsub(e) approx.= 2 x 10 13 cm -3 , Tsub(e)(O) approx.= 1.5 keV, Tsub(i)(O) approx.= 1.5 keV and Zsub(eff) approx.= 3. The preliminary results suggest a size-cubed scaling from PLT, and are consistent with Alcator C scaling where tau approx. nR 2 a. Initial measurements of plasma disruption characteristics indicate current decay rates of approx. 800 kA in 8 ms which is within the TFTR design requirement of 3 MA in 3 ms. (author)

  8. TFTR initial operations

    International Nuclear Information System (INIS)

    Young, K.M.; Bell, M.; Blanchard, W.R.

    1983-11-01

    The Tokamak Fusion Test Reactor (TFTR) has operated since December 1982 with ohmically heated plasmas. Routine operation with feedback control of plasma current, position, and density has been obtained for plasmas with I/sub p/ approx. = 800 kA, a = 68 cm, R = 250 cm, and B/sub t/ = 27 kG. A maximum plasma current of 1 MA was achieved with q approx. = 2.5. Energy confinement times of approx. 150 msec were measured for hydrogen and deuterium plasmas with anti n/sub e/ approx. = 2 x 10 13 cm -3 , T/sub e/ (0) approx. = 1.5 keV, T/sub i/ (0) approx. = 1.5 keV, and Z/sub eff/ approx. = 3. The preliminary results suggest a size-cubed scaling from PLT and are consistent with Alcator C scaling where tau approx. nR 2 a. Initial measurements of plasma disruption characteristics indicate current decay rates of approx. 800 kA in 8 ms which is within the TFTR design requirement of 3 MA in 3 ms

  9. TFTR data management system

    International Nuclear Information System (INIS)

    Randerson, L.; Chu, J.; Ludescher, C.; Malsbury, J.; Stark, W.

    1986-01-01

    Developments in the tokamak fusion test reactor (TFTR) data-management system supporting data acquisition and off-line physics data reduction are described. Data from monitor points, timing channels, transient recorder channels, and other devices are acquired and stored for use by on-line tasks. Files are transferred off line automatically. A configuration utility determines data acquired and files transferred. An event system driven by file arrival activates off-line reduction processes. A post-run process transfers files not shipped during runs. Files are archived to tape and are retrievable by digraph and shot number. Automatic skimming based on most recent access, file type, shot numbers, and user-set protections maintains the files required for post-run data reduction

  10. TFTR data management system

    International Nuclear Information System (INIS)

    Randerson, L.; Chu, J.; Ludescher, C.; Malsbury, J.; Stark, W.

    1986-01-01

    Developments in the tokamak fusion test reactor (TFTR) data management system supporting data management system supporting data acquisition and off-line physics data reduction are described. Data from monitor points, timing channels, and transient recorder channels and other devices are acquired and stored for use by on-line tasks. Files are transferred off-line automatically. A configuration utility determines data acquired and files transferred. An event system driven by file arrival activates off-line reduction processes. A post-run process transfers files not shipped during runs. Files are archived to tape and are retrievable by digraph and shot number. Automatic skimming based on most recent access, file type, shot numbers, and user-set protection maintains the files required for post-run data reduction

  11. TFTR toroidal field coil design

    International Nuclear Information System (INIS)

    Smith, G.E.; Punchard, W.F.B.

    1977-01-01

    The design of the Tokamak Fusion Test Reactor (TFTR) Toroidal Field (TF) magnetic coils is described. The TF coil is a 44-turn, spiral-wound, two-pancake, water-cooled configuration which, at a coil current of 73.3 kiloamperes, produces a 5.2-Tesla field at a major radius of 2.48 meters. The magnetic coils are installed in titanium cases, which transmit the loads generated in the coils to the adjacent supporting structure. The TFTR utilizes 20 of these coils, positioned radially at 18 0 intervals, to provide the required toroidal field. Because it is very highly loaded and subject to tight volume constraints within the machine, the coil presents unique design problems. The TF coil requirements are summarized, the coil configuration is described, and the problems highlighted which have been encountered thus far in the coil design effort, together with the development tests which have been undertaken to verify the design

  12. Tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

    Gouge, M.J.; Baylor, L.R.; Cole, M.J.; Combs, S.K.; Dyer, G.R.; Fehling, D.T.; Fisher, P.W.; Foust, C.R.; Langley, R.A.; Milora, S.L.; Qualls, A.L.; Wilgen, J.B.; Schmidt, G.L.; Barnes, G.W.; Persing, R.G.

    1992-01-01

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) phase. The existing TFTR deuterium pellet injector (DPI) has been modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed to provide pellets ranging from 3.3 to 4.5 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller. The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed, and the TPI was tested at ORNL with deuterium pellet. Results of the limited testing program at ORNL are described. The TPI is being installed on TFTR to support the D-D run period in 1992. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and secondary tritium containment systems and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  13. The tritium operations experience on TFTR

    International Nuclear Information System (INIS)

    Halle, A. von; Anderson, J.L.; Gentile, C.; Grisham, L.; Hosea, J.; Kamperschroer, J.; LaMarche, P.; Oldaker, M.; Nagy, A.; Raftopoulos, S.; Stevenson, T.

    1995-01-01

    The Tokamak Fusion Test Reactor (TFTR) tritium gas system is administratively limited to 5 grams of tritium and provides the feedstock gas for the neutral beam and torus injection systems. Tritium operations on TFTR began with leak checking of gas handling systems, qualification of the gas injection systems, and high power plasma operations using trace amounts of tritium in deuterium feedstock gas. Full tritium operation commenced with four highly diagnosed neutral beam pulses into a beamline calorimeter to verify planned tritium beam operating routines and to demonstrate the deuterium to tritium beam isotope exchange. Since that time, TFTR has successfully operated each of the twelve neutral beam ion sources in tritium during hundreds of tritium beam pulses and torus gas injections. This paper describes the TFTR tritium gas handling systems and TFTR tritium operations of the gas injection systems and the neutral beam ion sources. Tritium accounting and accountability is discussed, including tritium retention issues of the torus limiters and beam impinged surfaces of the beamline components. Also included is tritium beam velocity analysis that compares the neutral beam extracted ion species composition for deuterium and tritium and that determines the extent of beam isotope exchange on subsequent deuterium and tritium beam pulses. The required modifications to TFTR operating routines to meet the U.S. Department of Energy regulations for a low hazard nuclear facility and the problems encountered during initial tritium operations are described. (orig.)

  14. The tritium operations experience on TFTR

    International Nuclear Information System (INIS)

    von Halle, A.; Gentile, C.

    1994-01-01

    The Tokamak Fusion Test Reactor (TFTR) tritium gas system is administratively limited to 5 grains of tritium and provides the feedstock gas for the neutral beam and torus injection systems. Tritium operations on TFTR began with leak checking of gas handling systems, qualification of the gas injection systems, and high power plasma operations using using trace amounts of tritium in deuterium feedstock gas. Full tritium operation commenced with four highly diagnosed neutral beam pulses into a beamline calorimeter to verify planned tritium beam operating routines and to demonstrate the deuterium to tritium beam isotope exchange. Since that time, TFTR has successfully operated each of the twelve neutral beam ion sources in tritium during hundreds of tritium beam pulses and torus gas injections. This paper describes- the TFTR tritium gas handling systems and TFTR tritium operations of the gas injection systems and the neutral beam ion sources. Tritium accounting and accountability is discussed, including tritium retention issues of the torus limiters and beam impinged surfaces of the beamline components. Also included is tritium beam velocity analysis that compares the neutral beam extracted ion species composition for deuterium and tritium and that determines the extent of beam isotope exchange on subsequent deuterium and tritium beam pulses. The required modifications to TFTR operating routines to meet the US Department of Energy regulations for a low hazard nuclear facility and the problems encountered during initial tritium operations are described

  15. The roles of electric field shear and Shafranov shift in sustaining high confinement in enhanced reversed shear plasmas on the TFTR tokamak

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.A.

    1997-02-01

    The relaxation of core transport barriers in TFTR Enhanced Reversed Shear plasmas has been studied by varying the radial electric field using different applied torques from neutral beam injection. Transport rates and fluctuations remain low over a wide range of radial electric field shear, but increase when the local E x B shearing rates are driven below a threshold comparable to the fastest linear growth rates of the dominant instabilities. Shafranov-shift-induced stabilization alone is not able to sustain enhanced confinement

  16. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

  17. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

    This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

  18. Fast current ramp experiments on TFTR

    International Nuclear Information System (INIS)

    Fredrickson, E.D.; McGuire, K.; Goldston, R.J.

    1987-05-01

    Electron heat transport on TFTR and other tokamaks is several orders of magnitude larger than neoclassical calculations would predict. Despite considerable effort, there is still no clear theoretical understanding of this anomalous transport. The electron temperature profile T/sub e/(r), shape has shown a marked consistency on many machines, including TFTR, for a wide range of plasma parameters and heating profiles. This could be an important clue as to the process responsible for this enhanced thermal transport. In this paper 'profile consistency' in TFTR is described and an experiment which uses a fast current ramp to transiently decouple the current density profile J(r), and the T/sub e/(r) profiles is discussed. From this experiment the influence of J(r) on electron temperature profile consistency can be determined

  19. TFTR materials issues and problems during design and construction

    International Nuclear Information System (INIS)

    Sabado, M.; Little, R.

    1984-01-01

    TFTR as well as its contemporaries, T15, JT60, and JET, have important contributions to make towards our understanding of plasma conditions in the thermonuclear regime. One of the main objectives of TFTR is to produce fusion power densities approaching those in a fusion reactor, approx.= 1 Wcm -3 at Q approx.= 1-2. TFTR will be the first tokamak to routinely use deuterium tritium, and produce approx.= 10 19 fusion neutrons per pulse. With startup of TFTR on December 24, 1982, the demonstration of physics feasibility of 'breakeven' is close at hand. Since TFTR performance will be reactor relevant, the capability of materials/components to withstand the hostile effects of a plasma environment will be presented. It is intended that designers of future fusion devices benefit from the materials technology developments and applications on TFTR. In an attempt to comply with this mandate, this paper will describe TFTR issues on materials, their developments, selections, problems, and solutions. Special emphasis will be given, in particular, to the impurity control devices in TFTR, namely, the limiter and surface pumping system located inside the vacuum vessel. The plasma will interact with these components and they will be subjected to disruptions, a vacuum of 10 -6 to 10 -8 torr and a nominal temperatures of 0 C. 'Painful' materials development problems encountered will be reviewed, as well as important 'lessons learned'. A briefing on the materials of construction will be given, with some comments on the problems that developed and their solutions. (orig.)

  20. X-ray diagnostics for TFTR

    International Nuclear Information System (INIS)

    von Goeler, S.; Hill, K.W.; Bitter, M.

    1982-12-01

    A short description of the x-ray diagnostic preparation for the TFTR tokamak is given. The x-ray equipment consists of the limiter x-ray monitoring system, the soft x-ray pulse-height-analysis-system, the soft x-ray imaging system and the x-ray crystal spectrometer. Particular attention is given to the radiation protection of the x-ray systems from the neutron environment

  1. Pneumatic pellet injectors for TFTR and JET

    International Nuclear Information System (INIS)

    Combs, S.K.; Milora, S.L.

    1986-01-01

    This paper describes the development of pneumatic hydrogen pellet injectors for plasma fueling applications on the Tokamak Fusion Test Reactor (TFTR) and the Joint European Torus (JET). The performance parameters of these injectors represent an extension of previous experience and include pellet sizes in the range 2-6 mm in diameter and speeds approaching 2 km/s. Design features and operating characteristics of these pneumatic injectors are presented

  2. Operations analysis of the unscheduled summer machine opening of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory

    International Nuclear Information System (INIS)

    Viola, M.E.; McCann, J.

    1985-01-01

    During experimental operation, a problem developed with the mechanical integrity of the TFTR surface pumping system neutralizer plates that required a vacuum vessel entry for repairs. This problem, coupled with several less significant machine internal problems that had been developing, forced the decision to make an unscheduled vacuum vessel entry. An extended machine outage at that time would have had a severe impact on the experimental schedule. Therefore, the goal was to make repairs and return the vacuum vessel to a clean condition as quickly as possible. The total time required between the end of regularly scheduled activity and restoration of the machine capability to routinely obtain 1 MA disruption-free plasma was 12 days

  3. High beta and confinement studies on TFTR

    International Nuclear Information System (INIS)

    Navratil, G.A.; Bhattacharjee, A.; Iacono, R.; Mauel, M.E.; Sabbagh, S.A.; Kesner, J.

    1992-01-01

    A new regime of high poloidal beta operation in TFTR was developed in the course of the first two years of this project (9/25/89 to 9/24/91). Our proposal to continue this successful collaboration between Columbia University and the Massachusetts Institute of Technology with the Princeton Plasma Physics Laboratory for a three year period (9/25/91 to 9/24/94) to continue to investigate improved confinement and tokamak performance in high poloidal beta plasmas in TFTR through the DT phase of operation was approved by the DOE and this is a report of our progress during the first 9 month budget period of the three year grant (9/25/91 to 6/24/92). During the approved three year project period we plan to (1) extend and apply the low current, high QDD discharges to the operation of TFTR using Deuterium and Tritium plasma; (2) continue the analysis and plan experiments on high poloidal beta phenomena in TFTR including: stability properties, enhanced global confinement, local transport, bootstrap current, and divertor formation; (3) plan and carry out experiments on TFTR which attempt to elevate the central q to values > 2 where entry to the second stability regime is predicted to occur; and (4) collaborate on high beta experiments using bean-shaped plasmas with a stabilizing conducting shell in PBX-M. In the seven month period covered by this report we have made progress in each of these four areas through the submission of 4 TFTR Experimental Proposals and the partial execution of 3 of these using a total of 4.5 run days during the August 1991 to February 1992 run

  4. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

    Jassby, D.L.; Leinoff, S.

    1979-12-01

    A conceptual design has been carried out for an Engineering Test Station (ETS) which will provide structural support and utilities/instrumentation services for blanket modules positioned adjacent to the vacuum vessel of the TFTR (Tokamak Fusion Test Reactor). The ETS is supported independently from the Test Cell floor. The ETS module support platform is constructed of fiberglass to eliminate electromagnetic interaction with the pulsed tokamak fields. The ETS can hold blanket modules with dimensions up to 78 cm in width, 85 cm in height, and 105 cm in depth, and with a weight up to 4000 kg. Interfaces for all utility and instrumentation requirements are made via a shield plug in the TFTR igloo shielding. The modules are readily installed or removed by means of TFTR remote handling equipment

  5. TFTR CAMAC power supplies reliability

    International Nuclear Information System (INIS)

    Camp, R.A.; Bergin, W.

    1989-01-01

    Since the expected life of the Tokamak Fusion Test Reactor (TFTR) has been extended into the early 1990's, the issues of equipment wear-out, when to refurbish/replace, and the costs associated with these decisions, must be faced. The management of the maintenance of the TFTR Central Instrumentation, Control and Data Acquisition System (CICADA) power supplies within the CAMAC network is a case study of a set of systems to monitor repairable systems reliability, costs, and results of action. The CAMAC network is composed of approximately 500 racks, each with its own power supply. By using a simple reliability estimator on a coarse time interval, in conjunction with determining the root cause of individual failures, a cost effective repair and maintenance program has been realized. This paper describes the estimator, some of the specific causes for recurring failures and their correction, and the subsequent effects on the reliability estimator. By extension of this program the authors can assess the continued viability of CAMAC power supplies into the future, predicting wear-out and developing cost effective refurbishment/replacement policies. 4 refs., 3 figs., 1 tab

  6. Engineering analysis of TFTR disruption

    International Nuclear Information System (INIS)

    Murray, J.G.; Rothe, K.E.; Bronner, G.

    1984-09-01

    This report covers an engineering approach quantifying the currents, forces, and times, as well as plasma position, for the worst-case disruption based on engineerign circuit assumptions for the plasma. As the plasma moves toward the wall during the current-decay phase of disruption, the wall currents affect the rate of movement and, hence, the decay time. The calculated structure-induced currents differ considerably from those calculated using a presently available criterion, which specifies that the plasma remains stationary in the center of the torus while decaying in 10 ms. This report outlines the method and basis for the engineering calculation used to determine the current and forces as a function of the circuit characteristics. It provides specific calculations for the Tokamak Fusion Test Reactor (TFTR) with variations in parameters such as the thermal decay time, the torus resistance, and plasma temperature during the current decay. The study reviews possible ways to reduce the disruption damage of TFTR by reducing the magnitude of the plasma external field energy that is absorbed by the plasma during the current decay

  7. Engineering analysis of TFTR disruption

    Energy Technology Data Exchange (ETDEWEB)

    Murray, J.G.; Rothe, K.E.; Bronner, G.

    1984-09-01

    This report covers an engineering approach quantifying the currents, forces, and times, as well as plasma position, for the worst-case disruption based on engineerign circuit assumptions for the plasma. As the plasma moves toward the wall during the current-decay phase of disruption, the wall currents affect the rate of movement and, hence, the decay time. The calculated structure-induced currents differ considerably from those calculated using a presently available criterion, which specifies that the plasma remains stationary in the center of the torus while decaying in 10 ms. This report outlines the method and basis for the engineering calculation used to determine the current and forces as a function of the circuit characteristics. It provides specific calculations for the Tokamak Fusion Test Reactor (TFTR) with variations in parameters such as the thermal decay time, the torus resistance, and plasma temperature during the current decay. The study reviews possible ways to reduce the disruption damage of TFTR by reducing the magnitude of the plasma external field energy that is absorbed by the plasma during the current decay.

  8. [Analysis of momentum and impurity confinment in TFTR (1990)

    International Nuclear Information System (INIS)

    1990-01-01

    Work during the present grant period has been concentrated in two areas and are discussed in this report: (1) a review of momentum confinement experiments in tokamaks, of momentum confinement theories and of previous comparisons of the two; and (2) analysis and documentation of the dedicated power-scan rotation experiment performed on TFTR in September 1988

  9. Development of large insulator rings for the TOKAMAK Fusion Test Reactor

    International Nuclear Information System (INIS)

    Brown, T.; Tobin, A.

    1977-01-01

    Research and development leading to the manufacture of large ceramic insulator rings for the TFTR (TOKAMAK Fusion Test Reactor). Material applictions, fabrication approach and testing activities are highlighted

  10. TFTR D-T results

    International Nuclear Information System (INIS)

    Meade, D.M.

    1994-01-01

    Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of ∼ 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of ∼1 and yielded a maximum fusion power of ∼ 9.2 MW. The fusion power density in the core of the plasma was ∼ 1.8 MW m -3 approximating that expected in a D-T fusion reactor. A TFTR plasma with T/D density ratio of ∼ 1 was found to have ∼ 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of τ E ∼ A 0.6 . The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha particles. The ∼ 5% loss of alpha particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined energetic alphas and the resultant alpha ash density. At fusion power levels of 7.5 MW, fluctuations at the Toroidal Alfven Eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed

  11. Tritium pellet injection sequences for TFTR

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Milora, S.L.; Attenberger, S.E.; Singer, C.E.; Schmidt, G.L.

    1983-01-01

    Tritium pellet injection into neutral deuterium, beam heated deuterium plasmas in the Tokamak Fusion Test Reactor (TFTR) is shown to be an attractive means of (1) minimizing tritium use per tritium discharge and over a sequence of tritium discharges; (2) greatly reducing the tritium load in the walls, limiters, getters, and cryopanels; (3) maintaining or improving instantaneous neutron production (Q); (4) reducing or eliminating deuterium-tritium (D-T) neutron production in non-optimized discharges; and (5) generally adding flexibility to the experimental sequences leading to optimal Q operation. Transport analyses of both compression and full-bore TFTR plasmas are used to support the above observations and to provide the basis for a proposed eight-pellet gas gun injector for the 1986 tritium experiments

  12. Parametric variations of ion transport in TFTR

    International Nuclear Information System (INIS)

    Scott, S.D.; Ernst, D.

    1993-01-01

    This paper is divided into three roughly independent sections. The first is a historical review of the twenty year history of experimental ion heat transport measurements from many tokamaks. The second is a study of ion heat transport in Ohmic TFTR plasmas which shows that χi ∼ χe ∼ 15χi neo . Thus, ion heat transport is demonstrated to be strongly anomalous even the absence of auxiliary heating. The third section describes the variation of χi with local ion temperature in TFTR during auxiliary heating, with emphasis on characterizing the differecens between transport in the L-mode and supershot regimes. The results are consistent with the conjecture that improved ion energy confinement in supershot plasmas is caused by a high ratio of T 1 /T e

  13. TFTR neutral-beam test facility

    International Nuclear Information System (INIS)

    Turitzin, N.M.; Newman, R.A.

    1981-11-01

    TFTR Neutral Beam System will have thirteen discharge ion sources, each with its own power supply. Twelve of these will be utilized for supplemental heating of the TFTR tokamak plasma, while the thirteenth will be dedicated to an off-machine test chamber for source development and/or conditioning. A test installation for one source was set up using prototype equipment to discover and correct possible deficiencies, and to properly coordinate the equipment. This test facility represents the first opportunity for assembling an integrated system of hardware supplied by diverse vendors, each of whom designed and built his equipment to performance specifications. For the installation and coordination of the different portions of the total system, particular attention was given to personnel safety and safe equipment operation. This paper discusses various system components, their characteristics, interconnection and control. Results of the recently initiated test phase will be reported at a later date

  14. Segmentation strategies for the irradiated and tritium contaminated PPPL TFTR

    International Nuclear Information System (INIS)

    Walton, G.R.; Spampinato, P.T.

    1995-01-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory is scheduled to complete its final experiments in the Fall of 1995. As a result, the TFTR will be activated and tritium contaminated. After the experiments are complete, the TFTR will undergo Shutdown and Removal (S and R). The space vacated by the TFTR will be used for a new test reactor, the Tokamak Physics Experiment (TPX). Remote methods may be required to remove components and to segment the Vacuum Vessel. The TFTR has been studied to determine alternatives for the segmentation of the Vacuum Vessel from the inside (In-Vessel). The methodology to determine suitable strategies to segment the Vacuum Vessel from In-Vessel included several areas of concentration. These areas were segmentation locations, cutting/removal technologies, pros and cons, and cutting/removal technology delivery systems. The segmentation locations for easiest implementation and minimal steps in cutting and removal have been identified. Each of these will also achieve the baseline for packaging and shipment. The methods for cutting and removal of components were determined. In addition, the delivery systems were conceptualized

  15. Segmentation strategies for the irradiated and tritium contaminated PPPL TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Walton, G.R. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Litka, T.J. [Advanced Consulting Group, Inc., Chicago, IL (United States); Spampinato, P.T. [RHD Consultants, Inc., Princeton, NJ (United States)

    1995-02-09

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory is scheduled to complete its final experiments in the Fall of 1995. As a result, the TFTR will be activated and tritium contaminated. After the experiments are complete, the TFTR will undergo Shutdown and Removal (S and R). The space vacated by the TFTR will be used for a new test reactor, the Tokamak Physics Experiment (TPX). Remote methods may be required to remove components and to segment the Vacuum Vessel. The TFTR has been studied to determine alternatives for the segmentation of the Vacuum Vessel from the inside (In-Vessel). The methodology to determine suitable strategies to segment the Vacuum Vessel from In-Vessel included several areas of concentration. These areas were segmentation locations, cutting/removal technologies, pros and cons, and cutting/removal technology delivery systems. The segmentation locations for easiest implementation and minimal steps in cutting and removal have been identified. Each of these will also achieve the baseline for packaging and shipment. The methods for cutting and removal of components were determined. In addition, the delivery systems were conceptualized.

  16. Safety review and approval process for the TFTR

    International Nuclear Information System (INIS)

    Levine, J.D.; Howe, H.J.; Howe, K.E.

    1983-01-01

    The design, construction, and operation of the Tokamak Fusion Test Reactor (TFTR) has undergone an extensive safety and enviromental analysis involving Princeton Plasma Physics Laboratory (PPPL), the U.S. Department of Energy (DOE), the Ebasco/Grumman Industrial Subcontractor Team, and other organizations. This analysis, which is continuing during the TFTR operational phase, has been facilitated by the preparation, review and approval of several documents, including an Environmental Statement (Draft and Final), a Preliminary Safety Analysis Report (PSAR), a Final Safety Analysis Report (FSAR), Operations Safety Requirements (OSRs) and Safety Requirements (SRs), and various Operating and Maintenance Manuals. Through TFTR Safety Group participation in formal system design evaluations, change control boards, and reviews of project procurement and installation documentation, the TFTR Management Configuration Control System assures that all aspects of the project, including proposed design, installation and operational changes, receive prompt and thorough safety analyses. These efforts will continue as the TFTR Program moves into the neutral beam and D-T operational phases. The safety review and approval experience that has been acquired on the TFTR Project should serve as a foundation for similar efforts on future fusion devices

  17. TFTR diagnostic vacuum controller

    International Nuclear Information System (INIS)

    Olsen, D.; Persons, R.

    1981-01-01

    The TFTR diagnostic vacuum controller (DVC) provides in conjunction with the Central Instrumentation Control and Data Acquisition System (CICADA), control and monitoring for the pumps, valves and gauges associated with each individual diagnostic vacuum system. There will be approximately 50 systems on TFTR. Two standard versions of the controller (A and B) wil be provided in order to meet the requirements of two diagnostic manifold arrangements. All pump and valve sequencing, as well as protection features, will be implemented by the controller

  18. Neutron diagnostics on TFTR utilizing the Campbelling technique

    International Nuclear Information System (INIS)

    England, A.C.; Hendel, H.W.; Neischmidt, E.B.

    1986-01-01

    The authors report modified commercial neutron counting equipment installed on a tokamak fusion test reactor (TFTR) which utilizes the Campbelling theorem to monitor the neutron source strength at very high neutron count rates. Campbelling utilizes the large amplitude fluctuation from neutron events in the detectors to discriminate against small amplitude noise events. Source strengths yielding equivalent count rates a factor of five higher than possible in the conventional count rate mode have been obtained to date. The concept of Campbelling is discussed and the particular application to TFTR is illustrated

  19. Chromium getter studies in TFTR

    International Nuclear Information System (INIS)

    Dylla, H.F.; LaMarche, P.H.; Blanchard, W.R.

    1986-02-01

    We have studied the effects of the deposition of thin films (approx.0.1 μm) of chromium onto approx.70% of the torus area of the Tokamak Fusion Test Reactor (TFTR). The purpose of these experiments was to test the difference between high surface coverage and high pumping speed gettering schemes with respect to minimizing oxygen impurity generation in high power tokamak discharges. The initial Cr deposition had significant effects on vessel outgassing and subsequent plasma performance: the outgassing of H 2 O, CO, and CO 2 decreased by a factor of ten, oxygen impurity radiation decreased by a factor of two, the plasma Z/sub eff/ decreased from 1.3 to 1.1, and the plasma density limit increased by 20%. This improvement correlates with a significant reduction of the edge radiation as the density limit is approached. The effects of the initial and subsequent Cr depositions were relatively long lasting, exhibiting time constants of the order of weeks. We attribute the observed impurity reduction to a modification of the oxide surface on the vessel wall, which is apparently a significant impurity source for oxygen. 17 refs., 6 figs

  20. TFTR D and D Project: Final Examination and Testing of the TFTR TF-Coils

    International Nuclear Information System (INIS)

    Zatz, Irving J.

    2003-01-01

    In operation for nearly 15 years, TFTR (Tokamak Fusion Test Reactor) was not only a fusion science milestone, but a milestone of achievement in engineering as well. The TFTR DandD (Decommissioning and Decontamination) program provided a rare opportunity to examine machine components that had been exposed to a unique performance environment of greater than 100,000 mechanical and thermal load cycles. In particular, the possible examination of the TFTR toroidal-field (TF) coils, which met, then exceeded, the 5.2 Tesla magnetic field machine specification, could supply the answers to many questions that have been asked and debated since the coils were originally designed and built. A test program conducted in parallel with the DandD effort was the chance to look inside and examine, in detail, the TFTR TF coils for the first time since they were delivered encased to PPPL (Princeton Plasma Physics Laboratory). The results from such a program would provide data and insight that would not only be nefit PPPL and the fusion community, but the broader scientific community as well

  1. Tokamak experiments

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1987-01-01

    With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

  2. Nondimensional transport studies in TFTR

    International Nuclear Information System (INIS)

    Scott, S.D.; Mikkelsen, D.R.; Perkins, F.W.; Bell, M.G.; Bell, R.E.; Bush, C.E.; Fredrickson, E.D.; Grek, B.; Hill, K.W.; Janos, A.; Jobes, F.; Johnson, D.; Mansfield, D.K.; Owens, D.K.; Park, H.; Paul, S.; Ramsey, A.T.; Schivell, J.; Stratton, B.C.; Synakowski, E.J.; Tang, W.M.; Zarnstorff, M.C.; Ernst, D.

    1993-04-01

    The machine parameters (I p , P heat , R) required for ignition in ITER have generally been extrapolated from power-law regression fits to global τ E measurements on existing tokamaks. There remain important choices to be made in the form of the scaling relation which have not yet been resolved by theory. In particular, power flow Q(r) through a magnetic flux surface should scale as Q(r) = Q Bohm F where F = F(ρ*,β,ν*,s,T e /T i ,...) is a function of local, nondimensional plasma parameters and Q Bohm ∝ [n e T e 2 a/eB]. Projections to ITER can be reduced to establishing the dependence of F on ρ* = ρ i /a, because one can create plasmas in today's tokamaks which have similar values of the other nondimensional parameters. Two common scalings suggested by theory are Bohm (F independent of ρ*) and gyroBohm (F ∝ ρ*). Experiments have been carried out on TFTR to ascertain the dependence of F on ρ*, ν*, and β in L-mode plasmas, holding the other nondimensional parameters fixed. The observed variation of heat flow with ρ* was observed to be better described by Bohm scaling than gyroBohm. Comparisons with the critical gradient temperature transport model, which is gyroBohm in character, show that it overpredicts the temperature increase expected with increasing magnetic field. The ν* scan (remaining in the collisionless regime) revealed that the Bohm-normalized power flow is remarkably insensitive to collisionality, in agreement with ITER-P scaling. The β scan identified a deterioration of confinement with increasing β at fixed ρ* and ν*, of approximately the correct magnitude required to reconcile Bohm local transport scaling with ITER-P global scaling of τ E . This may suggest a role for electromagnetic phenomena in governing tokamak transport even at very low beta

  3. TFTR vertically viewing electron cyclotron emission diagnostic

    International Nuclear Information System (INIS)

    Taylor, G.

    1990-01-01

    The Tokamak Fusion Test Reactor (TFTR) Michelson interferometer has a spectral coverage of 75--540 GHz, allowing measurement of the first four electron cyclotron harmonics. Until recently the instrument has been configured to view the TFTR plasma on the horizontal midplane, primarily in order to measure the electron temperature profile. Electron cyclotron emission (ECE) extraordinary mode spectra from TFTR Supershot plasmas exhibit a pronounced, spectrally narrow feature below the second harmonic. A similar feature is seen with the ECE radiometer diagnostic below the electron cyclotron fundamental frequency in the ordinary mode. Analysis of the ECE spectra indicates the possibility of a non-Maxwellian 40--80 keV tail on the electron distribution in or near the core. During 1990 three vertical views with silicon carbide viewing targets will be installed to provide a direct measurement of the electron energy distribution at major radii of 2.54, 2.78, and 3.09 m with an energy resolution of approximately 20% at 100 keV. To provide the maximum flexibility, the optical components for the vertical views will be remotely controlled to allow the Michelson interferometer to be reconfigured to either the midplane horizontal view or one of the three vertical views between plasma shots

  4. Plasma wall interaction and tritium retention in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Amarescu, E.; Ascione, G.

    1996-01-01

    The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced an impressive number of record breaking results including core fusion power, ∼ 2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently, the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. The authors present recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues

  5. Plasma wall interaction and tritium retention in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Amarescu, E.; Ascione, G.

    1997-01-01

    The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced a number of record breaking results including core fusion power, ∝2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. Recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues are presented. (orig.)

  6. Measurement of Tritium Surface Distribution on TFTR Bumper Limiter Tiles

    International Nuclear Information System (INIS)

    Sugiyama, K.; Tanabe, T.; Skinner, C.H.; Gentile, C.A.

    2004-01-01

    The tritium surface distribution on graphite tiles used in the Tokamak Fusion Test Reactor (TFTR) bumper limiter and exposed to TFTR deuterium-tritium (D-T) discharges from 1993 to 1997 was measured by the Tritium Imaging Plate Technique (TIPT). The TFTR bumper limiter shows both re-/co-deposition and erosion. The tritium images for all tiles measured are strongly correlated with erosion and deposition patterns, and long-term tritium retention was found in the re-/co-depositions and flakes. The CFC tiles located at erosion dominated areas clearly showed their woven structure in their tritium images owing to different erosion yields between fibers and matrix. Significantly high tritium retention was observed on all sides of the erosion tiles, indicating carbon transport via repetition of local erosion/deposition cycles

  7. TFTR L mode energy confinement related to deuterium influx

    International Nuclear Information System (INIS)

    Strachan, J.D.

    1999-01-01

    Tokamak energy confinement scaling in TFTR L mode and supershot regimes is discussed. The main result is that TFTR L mode plasmas fit the supershot scaling law for energy confinement. In both regimes, plasma transport coefficients increased with increased edge deuterium influx. The common L mode confinement scaling law on TFTR is also inversely proportional to the volume of wall material that is heated to a high temperature, possibly the temperature at which the deuterium sorbed in the material becomes detrapped and highly mobile. The deuterium influx is increased by: (a) increased beam power due to a deeper heated depth in the edge components and (b) decreased plasma current due to an increased wetted area as governed by the empirically observed dependence of the SOL width upon plasma current. (author). Letter-to-the-editor

  8. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    Hendel, H.W.; Jassby, D.L.

    1989-11-01

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

  9. Heat pulse propagation studies in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Fredrickson, E.D.; Callen, J.D.; Colchin, R.J.; Efthimion, P.C.; Hill, K.W.; Izzo, R.; Mikkelsen, D.R.; Monticello, D.A.; McGuire, K.; Bell, J.D.

    1986-02-01

    The time scales for sawtooth repetition and heat pulse propagation are much longer (10's of msec) in the large tokamak TFTR than in previous, smaller tokamaks. This extended time scale coupled with more detailed diagnostics has led us to revisit the analysis of the heat pulse propagation as a method to determine the electron heat diffusivity, chi/sub e/, in the plasma. A combination of analytic and computer solutions of the electron heat diffusion equation are used to clarify previous work and develop new methods for determining chi/sub e/. Direct comparison of the predicted heat pulses with soft x-ray and ECE data indicates that the space-time evolution is diffusive. However, the chi/sub e/ determined from heat pulse propagation usually exceeds that determined from background plasma power balance considerations by a factor ranging from 2 to 10. Some hypotheses for resolving this discrepancy are discussed. 11 refs., 19 figs., 1 tab.

  10. Heat pulse propagation studies in TFTR

    International Nuclear Information System (INIS)

    Fredrickson, E.D.; Callen, J.D.; Colchin, R.J.

    1986-02-01

    The time scales for sawtooth repetition and heat pulse propagation are much longer (10's of msec) in the large tokamak TFTR than in previous, smaller tokamaks. This extended time scale coupled with more detailed diagnostics has led us to revisit the analysis of the heat pulse propagation as a method to determine the electron heat diffusivity, chi/sub e/, in the plasma. A combination of analytic and computer solutions of the electron heat diffusion equation are used to clarify previous work and develop new methods for determining chi/sub e/. Direct comparison of the predicted heat pulses with soft x-ray and ECE data indicates that the space-time evolution is diffusive. However, the chi/sub e/ determined from heat pulse propagation usually exceeds that determined from background plasma power balance considerations by a factor ranging from 2 to 10. Some hypotheses for resolving this discrepancy are discussed. 11 refs., 19 figs., 1 tab

  11. Upgrade to the Tritium Remote Control and Monitoring System for TFTR D and D

    International Nuclear Information System (INIS)

    Sichta, P.; Oliaro, G.; Sengupta, S.

    2002-01-01

    Since 1988, the Tritium Remote Control and Monitoring System (TRECAMS) has performed crucial functions in support of D-T [deuterium-tritium] operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory (PPPL). Although plasma operations on TFTR were completed in 1997, the need for TRECAMS continued. During this period TRECAMS supported the TFTR tritium systems, the TFTR's Shutdown and Safing phase, and the TFTR Decontamination and Decommissioning (D and D) project. The most critical function of the TRECAMS in the post-TFTR era has been to provide a real-time indication of the airborne tritium levels in the tritium areas and the (HVAC) stacks. TRECAMS is a critical tool in conducting safe TFTR D and D tritium-line breaks and other tritium-related work activities. Beginning in 1998, the failure rate of the system's hardware sharply increased. Furthermore, the specialized knowledge required to maintain the original software and hardware was diminishing. It soon became apparent that a failure of the TRECAMS could significantly impact the TFTR D and D project's cost and schedule. To preclude this, the TRECAMS hardware and software was upgraded in the year 2000 to use modern components. This paper will describe that successful upgrade, including a review of the engineering processes and our operating experiences with the upgraded system

  12. Demonstrating diamond wire cutting of the TFTR

    International Nuclear Information System (INIS)

    Rule, K.; Perry, E.; Larson, S.; Viola, M.

    2000-01-01

    The Tokamak Fusion Test Reactor (TFTR) ceased operation in April 1997 and decommissioning commenced in October 1999. The deuterium-tritium fusion experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the Tokamak (100 cubic meters) present a unique and challenging task for dismantling. Plasma arc cutting is the current baseline technology for the dismantlement of fission reactors. This technology is typically used because of its faster cutting times. Alternatively, an innovative approach for dismantlement of the TFTR is the use of diamond wire cutting technology. Recent improvements in diamond wire technology have allowed the cutting of carbon steel components such as pipe, plate, and tube bundles in heat exchangers. Some expected benefits of this technology include: significantly reduction in airborne contaminates, reduced personnel exposure, a reduced risk of spread of tritium contamination, and reduced overall costs as compared to using plasma arc cutting. This paper will provide detailed results of the diamond wire cutting demonstration that was completed in September of 1999, on a mock-up of this complex reactor. The results will identify cost, safety, industrial and engineering parameters, and the related performance of each situation

  13. Demonstrating diamond wire cutting of the TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Perry, E.; Larson, S.; Viola, M. [and others

    2000-02-24

    The Tokamak Fusion Test Reactor (TFTR) ceased operation in April 1997 and decommissioning commenced in October 1999. The deuterium-tritium fusion experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the Tokamak (100 cubic meters) present a unique and challenging task for dismantling. Plasma arc cutting is the current baseline technology for the dismantlement of fission reactors. This technology is typically used because of its faster cutting times. Alternatively, an innovative approach for dismantlement of the TFTR is the use of diamond wire cutting technology. Recent improvements in diamond wire technology have allowed the cutting of carbon steel components such as pipe, plate, and tube bundles in heat exchangers. Some expected benefits of this technology include: significantly reduction in airborne contaminates, reduced personnel exposure, a reduced risk of spread of tritium contamination, and reduced overall costs as compared to using plasma arc cutting. This paper will provide detailed results of the diamond wire cutting demonstration that was completed in September of 1999, on a mock-up of this complex reactor. The results will identify cost, safety, industrial and engineering parameters, and the related performance of each situation.

  14. Helium transport in TFTR

    International Nuclear Information System (INIS)

    Strachan, J.D.; Chan, A.

    1986-09-01

    Initial measurements of the 15 MeV protons produced in TFTR by the d( 3 He, p)α fusion reaction have been used to determine the time evolution of the central 3 He density. The signals following short 3 He gas puffs indicate inward transport times of about 100 msec

  15. Tokamak Fusion Test Reactor neutral beam injection system vacuum chamber

    International Nuclear Information System (INIS)

    Pedrotti, L.R.

    1977-01-01

    Most of the components of the Neutral Beam Lines of the Tokamak Fusion Test Reactor (TFTR) will be enclosed in a 50 cubic meter box-shaped vacuum chamber. The chamber will have a number of unorthodox features to accomodate both neutral beam and TFTR requirements. The design constraints, and the resulting chamber design, are presented

  16. Observations of Flaking of Co-deposited Layers in TFTR

    International Nuclear Information System (INIS)

    Gentile, C.A.; Skinner, C.H.; Young, K.M.

    1999-01-01

    Flaking of co-deposited layers in the Tokamak Fusion Test Reactor (TFTR) has been observed after the termination of plasma operations. This unexpected flaking affects approximately 15% of the tiles and appears on isotropic graphite tiles but not on carbon fiber composite tiles. Samples of tiles, flakes and dust were recently collected from the inside of the vacuum vessel and will be analyzed to better characterize the behavior of tritium on plasma facing components in DT fusion devices

  17. Thermostructural and mechanical aspects of the TFTR plasma limiter design

    International Nuclear Information System (INIS)

    Condolff, R.; Fixler, S.

    1977-01-01

    This paper presents the preliminary mechanical and thermostructural aspects of the TFTR (TOKAMAK Fusion Test Reactor) plasma limiter design. The evolution of the limiter design is traced through the various stages from conceptual design to the present state. Results of parametric limiter blade studies are presented. Design criteria, requirements, design loads (mechanical and thermal), material considerations, and remote handling problems are described. The design approach used to achieve a satisfactory plasma limiter and blade is discussed

  18. Thermostructural and mechanical aspects of the TFTR plasma limiter design

    International Nuclear Information System (INIS)

    Condolff, R.; Fixler, S.

    1978-01-01

    This paper presents the preliminary mechanical and thermostructural aspects of the TFTR (TOKAMAK Fusion Test Reactor) plasma limiter design. The evolution of the limiter design is traced through the various stages from conceptual design to the present state. Results of parametric limiter blade studies are presented. Design criteria, requirements, design loads (mechanical and thermal), material considerations, and remote handling problems are described. The design approach used to achieve a satisfactory plasma limiter and blade is discussed

  19. Four-channel ZnS scintillator measurements of escaping tritons in TFTR

    International Nuclear Information System (INIS)

    Zweben, S.J.

    1988-10-01

    A four-channel scintillation detector capable of measuring tritons, protons, and alphas escaping from a tokamak plasma was operated during the 1986 run period of the Tokamak Fusion Test Reactor (TFTR). Signals consistent with the expected 1 MeV triton behavior have been observed during deuterium operation. Backgrounds associated with neutrons, gammas, and soft x-rays have been evaluated in situ. Such a detector should be capable of measuring escaping alphas during the D/T phase of TFTR. 16 refs., 10 figs

  20. Four-channel ZnS scintillator measurements of escaping tritons in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Zweben, S.J.

    1988-10-01

    A four-channel scintillation detector capable of measuring tritons, protons, and alphas escaping from a tokamak plasma was operated during the 1986 run period of the Tokamak Fusion Test Reactor (TFTR). Signals consistent with the expected 1 MeV triton behavior have been observed during deuterium operation. Backgrounds associated with neutrons, gammas, and soft x-rays have been evaluated in situ. Such a detector should be capable of measuring escaping alphas during the D/T phase of TFTR. 16 refs., 10 figs.

  1. Coil protection calculator for TFTR

    International Nuclear Information System (INIS)

    Marsala, R.J.; Woolley, R.D.

    1987-01-01

    A new coil protection calculator (CPC) is presented in this paper. It is now being developed for TFTR's magnetic field coils will replace the existing coil fault detector. The existing fault detector sacrifices TFTR operating capability for simplicity. The new CPC will permit operation up to the actual coil limits by accurately and continuously computing coil parameters in real-time. The improvement will allow TFTR to operate with higher plasma currents and will permit the optimization of pulse repetition rates

  2. Design and analysis of the TFTR fixed limiters - 1

    International Nuclear Information System (INIS)

    Winkler, P.; Fixler, S.; Timlen, W.V.

    1981-01-01

    The operation of the Tokamak Fusion Test Reactor (TFTR) consists of two phases. In the first phase, the Tokamak systems will be tested and an ohmic heated plasma of 4 MW produced. The plasma limiter system for this phase consists of a set of movable and a set of fixed limiters. Because of the low power level during this phase, a design of passively cooled fixed limiters without tiles will satisfy the requirements. This limiter will be replaced by an actively cooled tile-covered axisymmetric limiter in the second phase. This paper discusses the design of the first phase fixed limiters only

  3. Impurity control in TFTR

    International Nuclear Information System (INIS)

    Cecchi, J.L.

    1980-06-01

    The control of impurities in TFTR will be a particularly difficult problem due to the large energy and particle fluxes expected in the device. As part of the TFTR Flexibility Modification (TEM) project, a program has been implemented to address this problem. Transport code simulations are used to infer an impurity limit criterion as a function of the impurity atomic number. The configurational designs of the limiters and associated protective plates are discussed along with the consideration of thermal and mechanical loads due to normal plasma operation, neutral beams, and plasma disruptions. A summary is given of the materials-related research, which has been a collaborative effort involving groups at Argonne National Laboratory, Sandia Laboratories, and Princeton Plasma Physics Laboratory. Conceptual designs are shown for getterng systems capable of regenerating absorbed tritium. Research on this topic by groups at the previously mentioned laboratories and SAES Research Laboratory is reviewed

  4. TFTR Motor Generator

    International Nuclear Information System (INIS)

    Murray, J.G.; Bronner, G.; Horton, M.

    1977-01-01

    A general description is given of 475 MVA pulsed motor generators for TFTR at Princeton Plasma Physics Laboratory. Two identical generators operating in parallel are capable of supplying 950 MVA for an equivalent square pulse of 6.77 seconds and 4,500 MJ at 0.7 power factor to provide the energy for the pulsed electrical coils and heating system for TFTR. The description includes the operational features of the 15,000 HP wound rotor motors driving each generator with its starting equipment and cycloconverter for controlling speed, power factor, and regulating line voltage during load pulsing where the generator speed changes from 87.5 to 60 Hz frequency variation to provide the 4,500 MJ or energy. The special design characteristics such as fatigue stress calculations for 10 6 cycles of operation, forcing factor on exciter to provide regulation, and low generator impedance are reviewed

  5. TFTR generator load assessment

    International Nuclear Information System (INIS)

    Heck, F.M.

    1975-10-01

    Typical experimental load demands on the TFTR generators are illustrated based on the electrical characteristics of the field coils, the coil leads, the main bus work, the various auxiliary bus work, the rectifiers, and transformers. The generator MW capacities are shown to be adequate for the proposed experimental operations with allowances made for variations in the final designs. The generator MVA capacities are shown to be adequate provided portions of the TF and EF rectifiers are freewheeled at selected times

  6. The TFTR maintenance manipulator

    International Nuclear Information System (INIS)

    Kungl, D.; Loesser, D.; Heitzenroeder, P.; Cerdan, G.

    1989-01-01

    TFTR plans to begin D-T experiments in mid 1990. The D-T experimental program will produce approximately one hundred shots, with a neutron generation rate of 10 19 neutrons per shot. This will result in high levels of activation in TFTR, especially in the vacuum vessel. The primary purpose of the Maintenance Manipulator is to provide a means of remotely performing certain defined maintenance and inspection tasks inside the vacuum torus so as to minimize personnel exposure to radiation. The manipulator consists of a six-link folding boom connected to a fixed boom on a movable carriage. The entire manipulator is housed in a vacuum antechamber connected to the vacuum torus, through a port formerly used for a vacuum pumping duct. The configuration extends 180 0 in either direction to provide complete coverage of the torus. The four 3500 l/s turbopumps which were formerly used in the pumping duct will be mounted on the antechamber. The manipulator will utilize two end effectors. The first, called a General Inspection Arm (GIA) provides a movable platform to an inspection camera and an in-vacuum leak detector. The second is a bilateral, force-reflecting pair of slave arms which utilize specially developed tools to perform several maintenance functions. All components except the slave arms are capable of operating in TFTR's vacuum environment and during 150 0 C bakeout of the torus. (orig.)

  7. Numerical Tokamak Project code comparison

    International Nuclear Information System (INIS)

    Waltz, R.E.; Cohen, B.I.; Beer, M.A.

    1994-01-01

    The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes

  8. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    Furth, H.P.

    1986-01-01

    The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)

  9. Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

    International Nuclear Information System (INIS)

    Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P

    2004-01-01

    Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values

  10. Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

    2004-09-01

    Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.

  11. The TFTR lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.; Bertone, P.C.; Creedon, R.L.; File, J.; Graumann, D.W.

    1985-01-01

    The Lithium Blanket Module (LBM) is an approximately 80X80X80 cm cubic module, representative of a helium-cooled lithium oxide fusion reactor blanket module, that will be installed on the TFTR (Tokamak Fusion Test Reactor) in late 1986. The principal objective of the LBM Program is to perform a series of neutron transport and tritium-breeding measurements throughout the LBM when it is exposed to the TFTR toroidal fusion neutron source, and to compare these data with the predictions of Monte Carlo (MCNP) neutronics codes. The LBM consists of 920 2.5-cm diameter breeder rods constructed of lithium oxide (Li 2 O) pellets housed in thin-walled stainless steel tubes. Procedures for mass-producing 25,000 Li 2 O pellets with satisfactory reproducibility were developed using purified Li 2 O powder, and fabrication of all the breeder rods was completed in early 1985. Tritium assay methods were investigated experimentally using both small lithium metal samples and LBM-type pellets. This work demonstrated that the thermal extraction method will be satisfactory for accurate evaluation of the minute concentrations of tritium expected in the LBM pellets (0.1-1nCi/g)

  12. Coherent and turbulent fluctuations in TFTR

    International Nuclear Information System (INIS)

    McGuire, K.; Arunasalam, V.; Bell, M.G.

    1987-04-01

    Classification of the sawteeth observed in the TFTR tokamak has been carried out to highlight the differences between the many types observed. Three types of sawteeth are discussed: ''simple,'' ''small,'' and ''compound.'' During the enhanced confinement discharges on TFTR, sawteeth related to q = 1 are usually not present, but a sawtooth-like event is sometimes observed. β approaches the Troyon limit only at low q/sub cyl/ with a clear reduction of achievable β/sub n/ at high q/sub cyl/. This suggests that a β/sub p/ limit, rather than the Troyon-Gruber limit, applies at high q/sub cyl/ in the enhanced confinement discharges. These discharges also reach the stability boundary for n → ∞ ideal MHD ballooning modes. Turbulence measurements in the scrape-off region with Langmuir and magnetic probes show strong edge density turbulence n/n = 0.3 - 0.5, with weak magnetic turbulence B/sub θ/B/sub θ/ > 5 x 10 -6 measured at the wall, but these measurements are very sensitive to local edge conditions

  13. Tritium environmental transport studies at TFTR

    International Nuclear Information System (INIS)

    Ritter, P.D.; Dolan, T.J.; Longhurst, G.R.

    1993-01-01

    Environmental tritium concentrations will be measured near the Tokamak Fusion Test Reactor (TFTR) to help validate dynamic models of tritium transport in the environment. For model validation the database must contain sequential measurements of tritium concentrations in key environmental compartments. Since complete containment of tritium is an operational goal, the supplementary monitoring program should be able to glean useful data from an unscheduled acute release. Portable air samplers will be used to take samples automatically every 4 hours for a weak after an acute release, thus obtaining the time resolution needed for code validation. Samples of soil, vegetation, and foodstuffs will be gathered daily at the same locations as the active air monitors. The database may help validate the plant/soil/air part of tritium transport models and enhance environmental tritium transport understanding for the International Thermonuclear Experimental Reactor (ITER)

  14. Tritium environmental transport studies at TFTR

    Science.gov (United States)

    Ritter, P. D.; Dolan, T. J.; Longhurst, G. R.

    1993-06-01

    Environmental tritium concentrations will be measured near the Tokamak Fusion Test Reactor (TFTR) to help validate dynamic models of tritium transport in the environment. For model validation the database must contain sequential measurements of tritium concentrations in key environmental compartments. Since complete containment of tritium is an operational goal, the supplementary monitoring program should be able to glean useful data from an unscheduled acute release. Portable air samplers will be used to take samples automatically every 4 hours for a week after an acute release, thus obtaining the time resolution needed for code validation. Samples of soil, vegetation, and foodstuffs will be gathered daily at the same locations as the active air monitors. The database may help validate the plant/soil/air part of tritium transport models and enhance environmental tritium transport understanding for the International Thermonuclear Experimental Reactor (ITER).

  15. Initial conditioning of the TFTR vacuum vessel

    International Nuclear Information System (INIS)

    Dylla, H.F.; Blanchard, W.R.; Krawchuk, R.B.; Hawryluk, R.J.; Owens, D.K.

    1984-01-01

    We report on the initial conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel prior to the initiation of first plasma discharges, and during subsequent operation with high power ohmically-heated plasmas. Following evacuation of the 86 m 3 vessel with the 10 4 1/s high vacuum pumping system, the vessel was conditioned by a 15 A dc glow discharge in H 2 at a pressure of 5 mTorr. Rapid-pulse discharge cleaning was used subsequently to preferentially condition the graphite plasma limiters. The effectiveness of the discharge cleaning was monitored by measuring the exhaust rates of the primary discharge products (CO/C 2 H 4 , CH 4 , and H 2 O). After 175 hours of glow discharge treatment, the equivalent of 50 monolayers of C and O was removed from the vessel, and the partial pressures of impurity gases were reduced to the range of 10 -9 -10 -10 Torr

  16. Remote leak detection for the TFTR

    International Nuclear Information System (INIS)

    Walthers, C.R.

    1977-01-01

    The planned design for the TFTR (TOKAMAK Fusion Test Reactor) remote leak detection system consists of a central console which controls the application of tracer gas to possible leak areas. Seals are tested by admitting tracer gas to machined cavities on the atmospheric side of the seal. The tracer gas is brought to the seal cavity by 1 / 8 -inch diameter tubes which connect to local tracer gas/vacuum manifolds located outside the protective radiation shielding. Vacuum shell walls and welds are checked by flowing tracer gas through annular heating/cooling passages. The detector will be either an MSLD (mass spectrometer leak detector) or an RGA (residual gas analyzer), the location of which is not finalized. Feasibility tests performed and planned include response and sensitivity measurements of possible tubing/detector configurations with several tracer gases

  17. Proposed TFTR electrical system

    International Nuclear Information System (INIS)

    Bronner, G.; Murray, J.

    1975-01-01

    The development of controlled thermonuclear fusion has progressed to the stage where the present facilities and energy available for future devices are not sufficient and must be increased by about a factor of ten. This report describes the proposed TFTR ac utility power distribution system, an energy storage motor generator flywheel facility, and the rectifier conversion equipment for the Toroidal Field Confining System (TF), Ohmic Heating System (OH), Equilibrium Field System (EF) and the Neutral Beam Heating System (NB). The general requirements are described and the special design considerations identified

  18. Probes for edge plasma studies of TFTR (invited)

    International Nuclear Information System (INIS)

    Manos, D.M.; Budny, R.V.; Kilpatrick, S.; Stangeby, P.; Zweben, S.

    1986-01-01

    Tokamak fusion test reactor (TFTR) probes are designed to study the interaction of the plasma with material surfaces such as the wall and limiters, and to study the transport of particles and energy between the core and edge. Present probe heads have evolved from prototypes in Princeton large torus (PLT), poloidal divertor experiment (PDX) [Princeton BETA experiment (PBX)], and the initial phase of TFTR operation. The newest heads are capable of making several simultaneous measurements and include Langmuir probes, heat flux probes, magnetic coils, rotating calorimeter fast ion probes, and sample exposure specimens. This paper describes these probe heads and presents some of the data they and their prototypes have acquired. The paper emphasizes measurement of transient plasma effects such as fast ion loss during auxiliary heating, the evolution of the edge plasma during heating, compression, and free expansion, and fluctuations in the edge plasma

  19. Neutron dosimetry for the TFTR Lithium-Blanket-Module program

    International Nuclear Information System (INIS)

    Harker, Y.D.; Tsang, F.Y.; Caffrey, A.J.; Homeyer, W.G.; Engholm, B.A.

    1981-01-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-a-kind neutronics experiment involving a prototypical fusion reactor blanket module with a distributed neutron source from the plasma of the TFTR at Princeton Plasma Physics Laboratory. The objectives of the LBM program are: (1) to test the capabilities of neutron transport codes when applied to fusion test reactor blanket conditions, and (2) to obtain tritium breeding performance data on a typical design concept of a fusion-reactor blanket. This paper addresses the issues relative to the measurement of neutron fields in the LBM, presents the results of preliminary design studies concerning neutron measurements and also presents the results of blanket mockup experiments performed at the Idaho National Engineering Laboratory

  20. DT simulation of ICRF heated supershots in TFTR using TRANSP

    International Nuclear Information System (INIS)

    Goldfinger, R.C.; Batchelor, D.B.; Phillips, C.K.; Budny, R.; Hammett, G.W.; Hosea, J.C.; McCune, D.M.; Stevens, J.E.; Wilson, J.R.

    1993-01-01

    The principal goal of ion cyclotron range of frequency (ICRF) heating on the Tokamak Fusion Test Reactor (TFTR) is to enhance plasma performance during the deuterium-tritium (DT) physics phase of operations. Strongly centralized ICRF heating may play a critical role in obtaining high Q DT and high β α operation in TFTR, as well as in future fusion reactors. ICRF heating of a dilute minority species leads to the formation of an energetic ion population that, in turn, provides strong central electron heating. The corresponding rise in the central electron temperature translates into an increase in the slowing-down time of either neutral beam or alpha particles in the discharge. Preliminary DT simulations of the experimental results in deuterium-deuterium (DD) plasmas performed with the TRANSP code are presented in this paper

  1. Neutron diagnostics on TFTR utilizing the Campbelling technique

    International Nuclear Information System (INIS)

    England, A.C.; Hendel, H.W.; Nieschmidt, E.B.

    1986-01-01

    Modified commercial equipment installed on the tokamak fusion test reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) utilizes Campbell's mean square voltage theorem to monitor the neutron source strength at neutron count rates orders of magnitude above the capability of the count rate mode. Campbelling uses the large amplitude fluctuations from neutron fission events in the detectors to discriminate against small amplitude γ ray and other noise events. Source strengths yielding equivalent count rates a factor of 5 greater than possible in the conventional count rate mode have been obtained to date. The concept of Campbelling is discussed and the particular application to TFTR is illustrated. Fundamental advantages are the extended useful range of the detectors by a factor of --10 4 and gamma rejection by a factor of --10 3 . Some results are shown and the neutron source strengths obtained are compared to those from conventional counting circuits and from other detectors whose outputs have not yet suffered counting losses

  2. Vacuum system transient simulator and its application to TFTR

    International Nuclear Information System (INIS)

    Sredniawski, J.

    1978-01-01

    The vacuum system transient simulator (VSTS) models transient gas transport throughout complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. VSTS is capable of treating gas models of up to 10 species, for all flow regimes from pure molecular to continuum. Viscous interactions between species are considered as well as non-uniform temperature of a system. Although this program was specifically developed for use on the Tokamak Fusion Test Reactor (TFTR) project at Princeton, it is a generalized tool capable of handling a broad range of vacuum system problems. During the TFTR engineering design phase, VSTS has been used in many applications. Two applications selected for presentation are: (1) torus vacuum pumping system performance between 400 Ci tritium pulses and (2) tritium backstreaming to neutral beams during pulses

  3. Vacuum system transient simulator and its application to TFTR

    International Nuclear Information System (INIS)

    Sredniawski, J.

    1977-01-01

    The vacuum system transient simulator (VSTS) models transient gas transport throughout complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. VSTS is capable of treating gas models of up to 10 species, for all flow regimes from pure molecular to continuum. Viscous interactions between species are considered as well as non-uniform temperature of a system. Although this program was specifically developed for use on the Tokamak Fusion Test Reactor (TFTR) project at Princeton, it is a generalized tool capable of handling a broad range of vacuum system problems. During the TFTR engineering design phase, VSTS has been used in many applications. Two applications selected for presentation are: torus vacuum pumping system performance between 400 Ci tritium pulses and tritium backstreaming to neutral beams during pulses

  4. Recent results from TFTR

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Bell, M.G.; Bitter, M.

    1984-05-01

    During the past year, the research activities on TFTR have encompassed three broad areas. The first was to extend the operating range of TFTR. Plasma currents up to 1.5 MA were achieved in discharges with a = 0.83 m, R = 2.55 m at a toroidal field of 2.7 T. In these large plasmas, the maximum line average density was 3.35 x 10 19 m -3 . The second activity was a study of the scaling of the energy confinement time, tau/sub E/, in ohmically heated discharges as a function of plasma current, density, and plasma size. These experiments indicate a favorable scaling of tau/sub E/ with size and density. Energy confinement times in excess of 0.25 s were obtained in deuterium discharges. The third activity was a study of adiabatic compression. During compression, the plasma current and ion temperature scaled approximately as predicted; however, the electron temperature and density scaled less strongly than predicted for ideal compression

  5. Long Term Tritium Trapping in TFTR and JET

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Young, K.M.; Coad, J.P.; Hogan, J.T.; Penzhorn, R.-D.; Bekris, N.

    2001-01-01

    Tritium retention in TFTR [Tokamak Fusion Test Reactor] and JET [Joint European Torus] shows striking similarities and contrasts. In TFTR, 5 g of tritium were injected into circular plasmas over a 3.5 year period, mostly by neutral-beam injection. In JET, 35 g were injected into divertor plasmas over a 6 month campaign, mostly by gas puffing. In TFTR, the bumper limiter provided a large source of eroded carbon and a major part of tritium was co-deposited on the limiter and vessel wall. Only a small area of the co-deposit flaked off. In JET, the wall is a net erosion area, and co-deposition occurs principally in shadowed parts of the inner divertor, with heavy flaking. In both machines, the initial tritium retention, after a change from deuterium [D] to tritium [T] gas puffing, is high and is due to isotope exchange with deuterium on plasma-facing surfaces (dynamic inventory). The contribution of co-deposition is lower but cumulative, and is revealed by including periods of D fueling that reversed the T/D isotope exchange. Ion beam analysis of flakes from TFTR showed an atomic D/C ratio of 0.13 on the plasma facing surface, 0.25 on the back surface and 0.11 in the bulk. Data from a JET divertor tile showed a larger D/C ratio with 46% C, 30% D, 20% H and 4% O. Deuterium, tritium, and beryllium profiles have been measured and show a thin less than 50 micron co-deposited layer. Flakes retrieved from the JET vacuum vessel exhibited a high tritium release rate of 2e10 Bq/month/g. BBQ modeling of the effect of lithium on retention in TFTR showed overlapping lithium and tritium implantation and a 1.3x increase in local T retention

  6. Feasability study of using the TFTR Thomson scattering system for q profile measurements

    International Nuclear Information System (INIS)

    Brizard, A.; Grewk, B.; Johnson, D.; LeBlanc, B.

    1986-01-01

    The results of a study made to determine the possibility of using the TFTR 76 channels Thomson scattering system to measure the direction of local magnetic fields in a tokamak plasma are presented. As this is a local measurement, this technique can in principle yield q profiles without the need of any de-convolution. The effect of the TFTR geometrical configuration and its various components on the expected measurement accuracy is discussed. The authors find that the measurement of q values within the inner half of the plasma should be possible, with only minor modification to the present TVTS system

  7. Fokker-Planck Modelling of Delayed Loss of Charged Fusion Products in TFTR

    International Nuclear Information System (INIS)

    Edenstrasser, J.W.; Goloborod'ko, V.Ya.; Reznik, S.N.; Yavorskij, V.A.; Zweben, S.

    1998-01-01

    The results of a Fokker-Planck simulation of the ripple-induced loss of charged fusion products in the Tokamak Fusion Test Reactor (TFTR) are presented. It is shown that the main features of the measured ''delayed loss'' of partially thermalized fusion products, such as the differences between deuterium-deuterium and deuterium-tritium discharges, the plasma current and major radius dependencies, etc., are in satisfactory agreement with the classical collisional ripple transport mechanism. The inclusion of the inward shift of the vacuum flux surfaces turns out to be necessary for an adequate and consistent explanation of the origin of the partially thermalized fusion product loss to the bottom of TFTR

  8. TFTR radiation contour and shielding efficiency measurements during D-D operations

    International Nuclear Information System (INIS)

    Kugel, H.W.; Ascione, G.; Elwood, S.; Gilbert, J.; Hwang, D.; Lewis, M.; Levine, J.; Ku, L.P.; Rule, K.; Hajnal, F.

    1994-11-01

    Extensive neutron and gamma radiation contour, shielding efficiency, and spectral measurements were performed during high power TFTR D-D operations at the tokamak Test Cell inner walls, ceiling, roof, and outer walls, in nearby control rooms, work areas, and personnel pathways, outdoors along the site fence at 125 m, and out to the nearest property lines at 180 m. The results confirmed that the efficiency of the basic radiation shielding was sufficient to allow the TFTR D-T experimental plan, and provide empirical guidance for simulating the radiation fields of future fusion reactors

  9. Discharge control and evolution in TFTR

    International Nuclear Information System (INIS)

    Mueller, D.; Bell, M.; Boody, F.; Bush, C.; Cecchi, J.L.; Davis, S.; Dylla, H.F.; Efthimion, P.C.

    1985-01-01

    The Tokamak Fusion Test Reactor (TFTR) was designed to explore plasma confinement and heating at reactor-like parameters. Operation of both the toroidal field and plasma current at full design parameters has been achieved and the plasma parameters are summarized in this work. Control of the discharge evolution has played an important role in attaining these parameters. The control of impurities in a tokamak is largely a result of the choice of limiter and wall materials, conditioning techniques and gettering. The impurity control procedures adopted during the run period ending April 13, 1985 are discussed. The discussion of discharge evolution and control is broken down into discharge initiation, volt-second consumption, current and density ramp-up and ramp-down. Also discussed is control of the current ramp-up using a plasma growing technique and the control of density using gas puffing, pellet injection and neutral beam fueling, along with a discussion of the density range which is found to increase plasma current

  10. Development of a maintenance manipulator for TFTR

    International Nuclear Information System (INIS)

    Holloway, C.

    1986-01-01

    The maintenance manipulator is a device permanently connected to the Tokamak Fusion Test Reactor (TFTR) vacuum vessel and is located in close proximity to the tokamak. It is used for the inspection and maintenance of in-vessel components whilst the machine remains under vacuum. The total system comprises a vacuum vessel ante-chamber that houses the manipulator, an articulated boom and carriage that transports and positions a dexterous end-effector, and end-effector that supports maintenance tooling, and an inspection system. Because of the maintenance manipulator's operating environment, there are many challenging engineering features, i.e., temperatures up to 150 0 C, changing magnetic fields in space and time that act on the manipulator whilst it is at rest, neutron neutron fluxes of up to 10/sup 11/cm/sup -2/s/sup -1/, and, last but not least, UHV conditions. This paper describes the development of the vacuum system, the maintenance manipulator, and inspective devices. It includes the methods employed to overcome the engineering difficulties and the application of information gained from other advanced technology programs, such as space and nuclear fission

  11. Transport of recycled deuterium to the plasma core in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Bell, M.G.; Budny, R.V.; Jassby, D.L.; Park, H.; Ramsey, A.T.; Stotler, D.P.; Strachan, J.D.

    1997-10-01

    The authors report a study of the fueling of the plasma core by recycling in the Tokamak Fusion Test Reactor (TFTR). They have analyzed discharges fueled by deuterium recycled from the limiter and tritium-only neutral beam injection. In these plasmas, the DT neutron rate provides a measure of the deuterium influx into the core plasma. They find a reduced influx with plasmas using lithium pellet conditioning and with plasmas of reduced major (and minor) radius. Modeling with the DEGAS neutrals code shows that the dependence on radius can be related to the penetration of neutrals through the scrape-off layer

  12. Development of large insulator rings for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Brown, T.; Tobin, A.

    1978-01-01

    This paper discusses research and development leading to the manufacture of large ceramic insulator rings for the TFTR (TOKAMAK Fusion Test Reactor). Material applications, fabrication approach and testing activities are highlighted

  13. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Perry, E.; Chrzanowski, J.; Gentile, C.; Parsells, R.; Rule, K.; Strykowsky, R.; Viola, M.

    2003-01-01

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

  14. High performance deuterium-tritium plasmas in TFTR

    International Nuclear Information System (INIS)

    Sabbagh, S.A.; Bell, M.G.

    1995-03-01

    Plasmas composed of nominally equal concentrations of deuterium and tritium (DT) have been created in TFTR with the goals of producing significant levels of fusion power and of examining the effects of DT fusion alpha particles. Conditioning of the limiter by the injection of lithium pellets has led to an approximate doubling of the energy confinement time, τ E , in supershot plasmas at high plasma current (I p ≤ 2.5 MA) and high heating power (P b ≤ 33 MW). Operation with DT typically results in an additional 20% increase in τ E . In the high poloidal beta, advanced tokamak regime in TFTR, confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.5 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasmas, exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Confinement of alpha particles appears to be classical and losses due to collective effects have not been observed. While small fluctuations in fusion product loss were observed during ELMs, no large loss was detected in DT plasmas

  15. ORNL compact loop antenna design for TFTR and Tore Supra

    International Nuclear Information System (INIS)

    Taylor, D.J.; Baity, F.W.; Bryan, W.E.; Hoffman, D.J.; McIlwain, R.L.; Ray, J.M.

    1987-01-01

    The goal supplemental ion cyclotron resonance heating (ICRH) of fusion plasma is to deliver power at high efficiencies deep within the plasma. The technology for fast-wave ICRH has reached the point of requiring ''proof-of-performance'' demonstration of specific antenna configurations of specific antenna configurations and their mechanical adequacy for operating in a fusion environment. Oak Ridge National Laboratory (ORNL) has developed the compact loop antenna concept based on a resonant double loop (RDL) configuration for use in both Tokamak Fusion Test Reactor (TFTR) and the Tore Supra ICRH programs. A description and a comparison of the technologies developed in the two designs are presented. The electrical circuit and the mechanical philosophy employed are the same for both antennas, but different operating environments result in substantial differences in the design of specific components. The ORNL TFTR antenna is designed to deliver 4 MW over a 2-s pulse, and the ORNL Tore Supra antenna is designed for 4 MW and essentially steady-state conditions. The TFTR design embodies the first operations compact RDL antenna, and the Tore Supra antenna extends the technology to an operational duty cycle consistent with reactor-relevant applications. 7 refs., 5 figs

  16. EMI free fiber optic strain sensor system for TFTR

    International Nuclear Information System (INIS)

    Szuchy, N.C.; Caserta, A.L.; Ferrara, A.A.; Squires, R.W.; Sredniawski, J.J.

    1983-01-01

    In certain applications, structural components are subjected to loadings in high electromagnetic interference (EMI) environments. The mechanical responses of these components must be monitored under rapidly varying electromagnetic fields. A Fiber Optic Strain Sensor System (FOSSS) is an acceptable solution since it is immune to EMI. Grumman Aerospace Corporation initiated the development of a FOSSS that can be used in high EMI situations where resistive/electronic-based strain measurement systems would not be effective, such as on the Tokamak Fusion Test Reactor (TFTR) during plasma disruption. Tests have indicated that because of their increased sensitivity due to the size of the fiber optic (FO) transducer (1-in. 2 ) and responsiveness due to the areal changes of the FO sensor, the strain tracking capability of FO sensors are excellent. For the TFTR application a jacketed 400-micron fiber capable of operating in a 250 0 C temperature environment was used. Continuous 30 foot lengths of high-temperature FO cables were affixed to 304 LN SS tabs, forming an integrated strain sensor and pigtail unit. By fusion splicing 400-micron room temperature fibers to the pigtails, the required runs (approximately 200 feet) to the TFTR data acquisition room were made with minimum coupling attenuation. Development methodology is discussed and test data presented

  17. Finite element modeling of TFTR poloidal field coils

    International Nuclear Information System (INIS)

    Baumgartner, J.A.; O'Toole, J.A.

    1986-01-01

    The Tokamak Fusion Test Reactor (TFTR) Poloidal Field (PF) coils were originally analyzed to TFTR design conditions. The coils have been reanalyzed by PPPL and Grumman to determine operating limits under as-built conditions. Critical stress levels, based upon data obtained from the reanalysis of each PF coil, are needed for input to the TFTR simulation code algorithms. The primary objective regarding structural integrity has been to ascertain the magnitude and location of critical internal stresses in each PF coil due to various combinations of electromagnetic and thermally induced loads. For each PF coil, a global finite element model (FEM) of a coil sector is being analyzed to obtain the basic coil internal loads and displacements. Subsequent fine mesh local models of the coil lead stem and lead spur regions produce the magnitudes and locations of peak stresses. Each copper turn and its surrounding insulation are modeled using solid finite elements. The corresponding electromagnetic and thermal analyses are similarly modeled. A series of test beams were developed to determine the best combination of MSC/NASTRAN-type finite elements for use in PF coil analysis. The results of this analysis compare favorably with those obtained by the earlier analysis which was limited in scope

  18. Meteorological data summaries for the TFTR from March 1984 to February 1985

    International Nuclear Information System (INIS)

    Kolibal, J.; Ku, L.P.; Liew, S.L.; Pierce, C.

    1985-06-01

    This report reviews the first year of meteorological data gathered for the Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) from March 1, 1984 to February 28, 1985. The meteorological station at TFTR is located at D-Site, to the east of the motor generator building as shown in Fig. 1. The station consists of a 60 m tower which is instrumented at 10, 30, and 60 m along with the associated equipment for data acquisition and logging. Instrumentation for the tower consists of measuring the temperature, wind speed, wind direction, dew point, and the standard deviation of the horizontal wind direction. The purpose of the station is to gather site specific meteorological data to assess atmospheric transport and dispersion for TFTR

  19. Evolution of the electron temperature profile of ohmically heated plasmas in TFTR

    International Nuclear Information System (INIS)

    Taylor, G.; Efthimion, P.C.; Arunasalam, V.

    1985-08-01

    Blackbody electron cyclotron emission was used to ascertain and study the evolution and behavior of the electron temperature profile in ohmically heated plasmas in the Tokamak Fusion Test Reactor (TFTR). The emission was measured with absolutely calibrated millimeter wavelength radiometers. The temperature profile normalized to the central temperature and minor radius is observed to broaden substantially with decreasing limiter safety factor q/sub a/, and is insensitive to the plasma minor radius. Sawtooth activity was seen in the core of most TFTR discharges and appeared to be associated with a flattening of the electron temperature profile within the plasma core where q less than or equal to 1. Two types of sawtooth behavior were identified in large TFTR plasmas (minor radius, a less than or equal to 0.8 m) : a typically 35 to 40 msec period ''normal'' sawtooth, and a ''compound'' sawtooth with 70 to 80 msec period

  20. 120-keV beam direct conversion system for TFTR injectors

    International Nuclear Information System (INIS)

    Hamilton, G.W.

    1976-01-01

    Several practical motivations exist for the development of beam direct conversion systems that are compatible with the injection systems of large experiments such as the Tokamak Fusion Test Reactor (TFTR). We present a preliminary design in which we analyze the most acute problems involved in scaling up existing designs and apparatus to fulfill TFTR requirements. Some of the questions addressed are the requirements for electron suppression, gas pumping, compactness, and power densities. A new idea is presented that allows for the handling of higher beam power. The gross savings in the capital cost of injector power supplies for the TFTR will be about $7.2 million, but the net savings will be somewhat less than this. This preliminary design has not yet revealed fundamental limitations with respect to the development of beam energy-recovery systems operating at high levels of current, voltage, and power densities

  1. Recent TFTR results

    International Nuclear Information System (INIS)

    Meade, D.M.; Arunasalam, V.; Bell, M.G.; Bell, R.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.; Cavallo, A.; Cheng, C.Z.; Chu, T.K.; Cohen, S.A.; Cowley, S.; Davis, S.L.; Dimock, D.L.; Efthimion, P.C.; Ehrhardt, A.B.; Fredrickson, E.; Furth, H.P.; Goldston, R.J.; Greene, G.; Grek, B.; Grisham, L.R.; Hammett, G.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.C.; Hulse, R.A.; Hsuan, H.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kaita, R.; Kaye, S.; Kieras-Phillips, C.; Kilpatrick, S.J.; Kugel, H.; La Marche, P.H.; LeBlanc, B.; Manos, D.M.; Mansfield, D.K.; Mazzucato, E.; McCarthy, M.P.; McCune, D.C.; McGuire, K.M.; Medley, S.S.; Mikkelsen, D.R.; Monticello, D.; Motley, R.; Mueller, D.; Murphy, J.; Nazikian, R.; Owens, D.K.; Park, H.; Park, W.; Paul, S.; Perkins, R.; Ramsey, A.T.; Redi, M.H.; Rewoldt, G.; Roquemore, A.L.; Rutherford, P.H.; Schilling, G.; Schivell, J.; Schmidt, G.L.; Scott, S.D.; Stevens, J.; Stodiek, W.; Stratton, B.C.; Synakowski, E.; Tang, W.A.; Taylor, G.; Timberlake, J.R.; Towner, H.H.; Ulrickson, M.; von Goeler, S.; Wieland, R.; Williams, M.D.; Wilson, J.R.; Wong, K.L.; Yamada, M.; Yoshikawa, S.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J.; Bush, C.E.; Dooling, J.; Dylla, H.F.; Fonck, R.J.; Roberts, D.; Howell, R.B.; Kesner, J.; Marmar, E.S.; Snipes, J.; Terry, J.L.; Nagayama, Y.; Pitcher, S.

    1991-07-01

    TFTR experiments have emphasized the optimization of high performance plasmas as well as studies of transport in high temperature plasmas. The recent installation of carbon composite tiles on the main bumper limiter has allowed operation with up to 32 MW of neutral beam injection without degradation of plasma performance by large bursts of carbon impurities (''carbon blooms''). Plasma parameters have been extended to T i (0) ∼ 35 keV, T e (0) ∼ 12 keV, n e (0) ∼1.2 x 10 20 m -3 producing D-D reaction rates of 8.8 x 10 16 reactions per second. The fusion parameter n e (0)τ E T i (0) in supershot plasmas is an increasing function of heating power up to an MHD stability limit, reaching values of ∼4.4 x 10 20 m -3 sec keV. Peaked-density-profile hot-ion plasmas with the edge characteristics of the H-mode have been produced in a circular cross-section limiter configuration with n e (0)τ E T i (0) values characteristic of supershots, namely up to four times those projected for standard H-modes with broad density profiles. Reduced transport is also observed in the core of high-density ICRF-heated plasmas when the density profile is peaked. At the highest performance, the central plasma pressure in TFTR reaches reactor level values of 6.5 atmospheres. In these regimes, MHD instabilities with m/n = 1/1, 2/1, 3/2 and 4/3 are often observed concurrent with a degradation in performance. High β p plasmas with var-epsilon β p ∼ 1.6 and β/(I/aB) ∼ 4.7 (%mT/MA) have demonstrated confinement enhancement over the low-mode confinement time with τ E /τ L ∼ 3.5 and a bootstrap current of about 65% of the total plasma current

  2. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1985-01-01

    The propagation of submillimeter-waves (smm) in tokamak plasmas has been investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses have been carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system has been employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes have been developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements

  3. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1986-01-01

    Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures

  4. Material selection for TFTR limiters

    International Nuclear Information System (INIS)

    Ulrickson, M.

    1980-10-01

    The requirements for the material to be used as the first surface of limiters in TFTR are that it: (1) withstand a heat flux of 1 kw/cm 2 for a pulse length of 1.5s and a duty cycle of 1/200 for 10 5 cycles, (2) withstand the thermal and electro-magnetic loads from 10 4 plasma current disruptions lasting about 200 μs, (3) generate impurities at a rate low enough to meet impurity control requirements (which depend on the atomic number of the material) for TFTR, and (4) have tritium retention characteristics consistent with tritium inventory requirements for TFTR. An extensive set of material tests using electron beams, neutral beams, and plasma bombardment have been carried out to identify materials which can meet the thermal requirements of the above

  5. TFTR CAMAC instrumentation system

    International Nuclear Information System (INIS)

    Del Gatto, H.J.; Bradish, C.J.

    1983-01-01

    The TFTR Central Instrumentation Control and Data Acquisition (CICADA) system makes extensive use of CAMAC equipment. The system consists of eight CAMAC highways operating from eight Gould 75/32 computers. Links up to 3.5 miles in length with more than fifty CAMAC crates have been implemented and are currently in use. Data transfer along the highway is implemented in bit serial format. The link speed is run at 5MHz. The length and complexity of the link requires the reformatting of the NRZ input/output format of the L-2 crate controller. U-Port adapter modules are used to interface the modified serial highway to the L-2 controllers. The modified serial highway uses a transmission technique that requires the distribution of both Bi-Phase encoded data and a 5MHz clock. The Serial Driver interfaces to the GOULD computer through use of a High Speed Data (HSD) interface board which attaches to the computers internal bus. All transfers to and from the computer are accomplished by direct memory access (DMA). In addition to the standard CAMAC link the system also includes a Block Transfer (BT) system. This system provides an alternate path for transferring data between the computers and the CAMAC modules. The BT system is interfaced to the host computers through HSD boards and to the CAMAC crates through use of an auxiliary crate controllers

  6. DATA management for TFTR

    International Nuclear Information System (INIS)

    Christianson, G.B.; Chu; Randerson, L.R.

    1983-01-01

    The TFTR experiment generates such a large amount of data each shot that only a restricted subset of the acquired data is displayed quickly, so that operational physicists can control and direct the experiments. The authors discuss the software tasks and data structures used to acquire this summary data and present it in a graphical form. After the summary data has been acquired and displayed, the large quantity of diagnostic physics and engineering data must be acquired and placed on bulk storage devices. They describe the software tasks and data bases used to accomplish this, and indicate future enhancements. Experimental systems often require peripheral data files and data bases. They describe the means by which auxiliary files are associated with primary acquired data for a shot, and discuss ancillary data bases. The authors outline future plans for auxiliary data base management to aid in the offline analysis of past data. Summary, raw and processed data must be reliably archived to a permanent storage medium. The archival procedure and management of the archival directory to permit orderly retrieval of data for offline analysis are described. Finally, archived data must be formatted in a standardized fashion to permit access by a broad community of users. At the same time, the large amounts of data to be archived make an efficient format necessary. The formatting of data files and outline future plans for transmittal of archived data to other computer systems are described

  7. Full power in the European tokamak

    International Nuclear Information System (INIS)

    Lallia, P.P.; Hugon, M.

    1987-01-01

    A new research campaign begins at Jet (Abingdon, UK). At this occasion, authors review and compare the performances of the three big Tokamaks that are currently in competition: Jet, JT60 and TFTR, insisting upon the European one. Conditions of ignition are reviewed together and energy losses are specified. Magnetic configurations used in tokamaks are shown, together with the technological responses brought these last years

  8. Modification and final alignment of the TFTR bumper limiter

    International Nuclear Information System (INIS)

    McSmith, M.D.; Loesser, G.D.; Owens, D.K.

    1994-01-01

    During the past three Tokamak Fusion Test Reactor (TFTR) vacuum vessel machine openings, an extensive effort was undertaken to optimize the distribution of heating of the bumper limiter tiles. The optimization was achieved by locating the limiter tiles relative to the toroidal magnetic field and adjusting their position relative to the magnetic field rather than to fixed points in the vacuum vessel walls. This paper will discuss the results of these alignments as measured during operation with the limiter thermocouple system and subsequent visual inspection during this past TFTR vacuum vessel opening. During the most recent in-vessel inspection (January 1993), damage to the top and bottom rows of the bumper limiter tiles was noted. More tiles were damaged on the lower row than the upper row. Tiles on the right side of the bottom row and to a lesser extent tiles on the left side of the top row were damaged. The location of the damage corresponds to the plasma power flux direction. Theories explaining the asymmetric damage (bottom versus top) are summarized. Princeton Plasma Physics Laboratory (PPL) began a program to replace 223 of the originally installed tiles made from POCO AFX-5Q graphite. Of these 223 tiles, 151 were replaced with tiles made from carbon-fiber-composite (CFC) and 158 of these tiles were re-designed for installation on the top or bottom rows. The re-designed tiles have a tapered edge that reduces the angle of incidence of the power flux on the edge surface that was over-heating. This paper will review the in-vessel work and discuss the final modification of the TFTR bumper limiter to alleviate further damage at these locations prior to DT operation of TFTR

  9. Neutron emission from TFTR supershots

    International Nuclear Information System (INIS)

    Strachan, J.D.; Bell, M.G.; Bitter, M.; Budny, R.; Hawryluk, R.; Hill, K.W.; Hsuan, H.; Jassby, D.L.; Johnson, L.C.; LeBlanc, B.; Mansfield, D.; Meade, D.; Mikkelsen, D.R.; Mueller, D.; Park, H.; Ramsey, A.; Scott, S.; Synakowski, E.; Taylor, G.; Marmer, E.; Snipes, J.; Terry, J.

    1992-10-01

    Empirical scaling relations are deduced describing the neutron emission from TFTR supershots using a data base that includes all of the supershot plasmas (525) from the 1990 campaign. A physics-based scaling for the neutron emission is derived from the dependence of the central plasma parameters on machine settings and the energy confinement time. This scaling has been used to project the fusion rate for equivalent DT plasmas in TFTR, and to explore machine operation space which optimizes the fusion rate. Increases in neutron emission are possible by either increasing the toroidal magnetic field or further improving the limiter conditioning

  10. Tritium experience in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

    1998-01-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

  11. Scope and status of the USA Engineering Test Facility including relevant TFTR research and development

    International Nuclear Information System (INIS)

    Becraft, W.R.; Reardon, P.J.

    1980-01-01

    The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The progress toward the design and construction of the ETF will reflect the significant achievements of past, present, and future experimental tokamak devices. Some of the features of this foundation of experimental results and relevant engineering designs and operation will derive from the Tokamak Fusion Test Reactor (TFTR) Project, now nearing the completion of its construction phase. The ETF would provide a test-bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy (OFE) established the ETF Design Center activity to prepare the design of the ETF. This paper describes the design status of the ETF and discusses some highlights of the TFTR R and D work

  12. Scope and status of the USA Engineering Test Facility including relevant TFTR research and development

    International Nuclear Information System (INIS)

    Becraft, W.R.; Reardon, P.J.

    1981-01-01

    The vehicle by which the fusion programme would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The progress toward the design and construction of the ETF will reflect the significant achievements of past, present, and future experimental tokamak devices. Some of the features of this foundation of experimental results and relevant engineering designs and operation will derive from the Tokamak Fusion Test Reactor (TFTR) Project, now nearing the completion of its construction phase. The ETF would provide a test-bed for reactor components in the fusion environment. To initiate preliminary planning for the ETF decision, the Office of Fusion Energy (OFE) established the ETF Design Center activity to prepare the design of the ETF. This paper describes the design status of the ETF and discusses some highlights of the TFTR R and D work. (author)

  13. Diagnostic interface problems on TFTR

    International Nuclear Information System (INIS)

    Goldfarb, S.

    1977-01-01

    Diagnostic equipment on TFTR has functional interfaces with many machine systems. Salient requirements include plasma access, environmental resistance to thermal, magnetic and radiation effects, automated data acquisition and controls, remote handling and personnel safety. Problems imposed by these requirements and the solutions being considered are described

  14. Limiter H-mode experiments on TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Bush, C [Oak Ridge National Lab., TN (USA); Bretz, N L; Fredrickson, E D; McGuire, K M; Nazikian, R; Park, H K; Schivell, J; Taylor, G; Bitter, B; Budny, R; Cohen, S A; Kilpatrick, S J; LeBlanc, B; Manos, D M; Meade, D; Paul, S F; Scott, S D; Stratton, B C; Synakowski, E J; Towner, H H; Weiland, R M; Arunasalam, V; Bateman, G; Bell, M G; Bell, R; Boivin, R; Cavallo, A; Cheng, C Z; Chu, T K; Cowl,

    1990-12-15

    Limiter H-modes with centrally peaked density profiles have been obtained in TFTR using a highly conditioned graphite limiter. The transition to these centrally peaked H-modes takes place from the supershot to the H-mode rather than the usual L- to H-mode transition observed on other tokamaks. Bi-directional beam heating is required to induce the transition. Density peaking factors, n{sub e}(0)/{l angle}n{sub e}{r angle}, >2.3 are obtained and at the same time the H-mode characteristics are similar to those of limiter H-modes on other tokamaks and the global confinement, {tau}{sub E}, can be >2.5 times L-mode scaling. The TRANSP analysis shows that transport in these H-modes is similar to that of supershots within the inner 60 cm of the plasma, but the stored electron energy (calculated using measured values of T{sub e} and n{sub e}) is higher for the H-mode at the plasma edge. Microwave scattering near the edge shows broad spectra at k = 5.5 cm{sup {minus}1} which begin at the drop in D{sub {alpha}} radiation and are strongly shifted in the electron diamagnetic drift direction. At the same time beam emission spectroscopy shows a coherent mode near the boundary with m = 15--20 at 20--30 kHz which is propagating in the ion direction. During an ELM event these apparent rotations cease and Mirnov fluctuations in the 50--500 kHz increase in intensity.

  15. Conceptual design of a neutral-beam injection system for the TFTR

    International Nuclear Information System (INIS)

    Ehlers, K.W.; Berkner, K.H.; Cooper, W.S.; Hooper, E.B.; Pyle, R.V.; Stearns, J.W.

    1975-11-01

    The neutral-beam injection requirements for heating and fueling the next generation of fusion reactor experiments far exceed those of present devices; the neutral-beam systems needed to meet these requirements will be large and complex. A conceptual design of a TFTR tokamak injection system to produce 120 keV deuterium-ion beams with a total power of about 80 MW is given

  16. Expansion of parameter space for Toroidal Alfven Eigenmode experiments in TFTR

    International Nuclear Information System (INIS)

    Wong, K.L.; Wilson, J.R.; Chang, Z.Y.; Fredrickson, E.; Hammett, G.W.; Bush, C.; Nazikian, R.; Phillips, C.K.; Snipes, J.; Taylor, G.

    1993-05-01

    Several techniques were used to excite toroidal Alfven Eigenmodes in the Tokamak Fusion Test Reactor (TFTR) at magnetic fields above 10 kG. These involve pellet injection to raise the plasma density, variation of plasma current to change the energetic ion orbit and the q-profile, and ICRF heating to produce energetic hydrogen ions at velocities comparable to 3.5 MeV alpha particles. These experimental results are presented and relevance to fusion reactors are discussed

  17. TFTR power conversion and plasma feedback systems

    International Nuclear Information System (INIS)

    Neumeyer, C.

    1985-01-01

    Major components of the Tokamak Fusion Test Reactor (TFTR) power conversion system include 39 thyristor rectifier power supplies, 12 energy storage capacitor banks, and 6 ohmic heating interrupters. These components are connected in various series/parallel configurations to provide controlled pulses of current to the Toroidal Field (TF), Ohmic Heating (OH), Equilibrium (vertical) Field (EF), and Horizontal Field (HF) magnet coil systems. Real-time control of the power conversion system is accomplished by a centralized dedicated computer; local control is minimal. Power supply firing angles, capacitor bank charge and discharge commands, interrupter commands, etc., are all determined and issued by the central computer. Plasma Position and Current Control (PPCC) reference signals to power conversion (OH, EF, HF) are determined by separate analog electronics but invoked through the power conversion computer. Real-time fault sensing of plasma parameters, gas injection, neutral beams, etc., are monitored by a separate Discharge Fault System (DFS) but routed through the power conversion computer for pre-programmed shutdown response

  18. High beta and confinement studies on TFTR

    International Nuclear Information System (INIS)

    1990-01-01

    In the first 8 months of this project we have made substantial progress toward the goals set out in our original proposal. Our plan to access new regimes of operation at high values of var-epsilon β p using low current discharges in TFTR has worked extremely well and a new regime of operation has indeed been found in the course of our execution of TFTR Experimental Proposal 146 which involved our operation of TFTR on 9 November 1989, 19--20 January 1990 and 1--2 February 1990. The status of our high var-epsilon β p work on TFTR is given and is extracted from our paper submitted for presentation to the 1990 EPS meeting in Amsterdam. We have also performed an analysis of the energetic particle stabilization requirements for TFTR Supershots, and developed methods for analysis and a theory of perturbative transport measurements in TFTR

  19. Results from D-T experiments on TFTR and implications for achieving an ignited plasma

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Blanchard, W.

    1998-01-01

    Progress in the performance of tokamak devices has enable not only the production of significant bursts of fusion energy from deuterium-tritium plasmas in the Tokamak Fusion Test Reactor (TFTR) and the Joint European Torus (JET) but, more importantly, the initial study of the physics of burning magnetically confined plasmas. As a result of the worldwide research on tokamaks, the scientific and technical issues for achieving an ignited plasma are better understood and the remaining questions more clearly defined. The principal research topics which have been studied on TFTR are transport, magnetohydrodynamic stability, and energetic particle confinement. The integration of separate solutions to problems in each of these research areas has also been of major interest. Although significant advances, such as the reduction of turbulent transport by means of internal transport barriers, identification of the theoretically predicted bootstrap current, and the study of the confinement of energetic fusion alpha-particles have been made, interesting and important scientific and technical issues remain. In this paper, the implications for the TFTR experiments for overcoming these remaining issues will be discussed

  20. Transport analysis of TFTR experiments

    International Nuclear Information System (INIS)

    Goldston, R.; McCune, D.; Zarnstorff, M.; Hammett, G.; Scott, S.

    1991-01-01

    The purpose of this investigation was to complete the analysis of TFTR data which was under progress. The main emphasis was to study the effects of heating profile and resulting density and temperature profiles on transport through the comparison between beam heated plasmas with hollow and centrally peaked heating profiles (edge vs. center heating). The analysis has been completed and a manuscript has been prepared for publication in Nuclear Fusion. A proposal to perform a similar experiment using ICRF heating to decouple heating profile effects from density profile effects was submitted and was approved by the TFTR. ICRF heating enables the heating profile and the power partition between ions and electrons to be controlled. The experiment was scheduled twice, but it had to be postponed both times

  1. TFTR neutral beam power system

    International Nuclear Information System (INIS)

    Deitz, A.; Murray, H.; Winje, R.

    1977-01-01

    The TFTR NB System will be composed of four beam lines, each containing three ion sources presently being developed for TFTR by the Lawrence Berkeley Laboratories (LBL). The Neutral Beam Power System (NBPS) will provide the necessary power required to operate these Ion Sources in both an experimental or operational mode as well as test mode. This paper describes the technical as well as the administrative/management aspects involved in the development and building of this system. The NBPS will combine the aspects of HV pulse (120 kV) and long pulse width (0.5 sec) together to produce a high power system that is unique in the Electrical Engineering field

  2. TFTR magnetic field design analyses

    International Nuclear Information System (INIS)

    Davies, K.; Iwinski, E.; McWhirter, J.M.

    1975-11-01

    The three main magnetic field windings for the TFTR are the toroidal field (TF) windings, the ohmic heating (OH) winding, and the equilibrium field (EF) winding. The following information is provided for these windings: (1) descriptions, (2) functions, (3) magnetic designs, e.g., number and location of turns, (4) design methods, and (5) descriptions of resulting magnetic fields. This report does not deal with the thermal, mechanical support, or construction details of the windings

  3. Central ignition scenarios for TFTR

    International Nuclear Information System (INIS)

    Zweben, S.J.; Redi, M.H.; Bateman, G.

    1986-03-01

    The possibility of obtaining ignition in TFTR by means of very centrally peaked density profiles is examined. It is shown that local central alpha heating can be made to exceed local central energy losses (''central ignition'') under global conditions for which Q greater than or equal to 1. Time dependent 1-D transport simulations show that the normal global ignition requirements are substantially relaxed for plasmas with peaked density profiles. 18 refs., 18 figs

  4. TFTR Experimental Data Analysis Collaboration

    International Nuclear Information System (INIS)

    Callen, J.D.

    1993-01-01

    The research performed under the second year of this three-year grant has concentrated on a few key TFTR experimental data analysis issues: MHD mode identification and effects on supershots; identification of new MHD modes; MHD mode theory-experiment comparisons; local electron heat transport inferred from impurity-induced cool pulses; and some other topics. Progress in these areas and activities undertaken in conjunction with this grant are summarized briefly in this report

  5. Coil protection calculator for TFTR

    International Nuclear Information System (INIS)

    Marsala, R.J.; Lawson, J.E.; Persing, R.G.; Senko, T.R.; Woolley, R.D.

    1989-01-01

    A new coil protection system (CPS) is being developed to replace the existing TFTR magnetic coil fault detector. The existing fault detector sacrifices TFTR operating capability for simplicity. The new CPS, when installed in October of 1988, will permit operation up to the actual coil stress limits parameters in real-time. The computation will be done in a microprocessor based Coil Protection Calculator (CPC) currently under construction at PPL. THe new CPC will allow TFTR to operate with higher plasma currents and will permit the optimization of pulse repetition rates. The CPC will provide real-time estimates of critical coil and bus temperatures and stresses based on real-time redundant measurements of coil currents, coil cooling water inlet temperature, and plasma current. The critical parameter calculations are compared to prespecified limits. If these limits are reached or exceeded, protection action will be initiated to a hard wired control system (HCS), which will shut down the power supplies. The CPC consists of a redundant VME based microprocessor system which will sample all input data and compute all stress quantities every ten milliseconds. Thermal calculations will be approximated every 10ms with an exact solution occurring every second. The CPC features continuous cross-checking of redundant input signal, automatic detection of internal failure modes, monitoring and recording of calculated results, and a quick, functional verification of performance via an internal test system. (author)

  6. TFTR plasma feedback systems

    International Nuclear Information System (INIS)

    Efthimion, P.; Hawryluk, R.J.; Hojsak, W.; Marsala, R.J.; Mueller, D.; Rauch, W.; Tait, G.D.; Taylor, G.; Thompson, M.

    1985-01-01

    The Tokamak Fusion Test Reactor employs feedback control systems for four plasma parameters, i.e. for plasma current, for plasma major radius, for plasma vertical position, and for plasma density. The plasma current is controlled by adjusting the rate of change of current in the Ohmic Heating (OH) coil system. Plasma current is continuously sensed by a Rogowski coil and its associated electronics; the error between it and a preprogrammed reference plasma current history is operated upon by a ''proportional-plusintegral-plus-derivative'' (PID) control algorithm and combined with various feedforward terms, to generate compensating commands to the phase-controlled thyristor rectifiers which drive current through the OH coils. The plasma position is controlled by adjusting the currents in Equilibrium Field and Horizontal Field coil systems, which respectively determine the vertical and radial external magnetic fields producing J X B forces on the plasma current. The plasma major radius position and vertical position, sensed by ''B /sub theta/ '' and ''B /sub rho/ '' magnetic flux pickup coils with their associated electronics, are controlled toward preprogrammed reference histories by allowing PID and feedforward control algorithms to generate commands to the EF and HF coil power supplies. Plasma density is controlled by adjusting the amount of gas injected into the vacuum vessel. Time-varying gains are used to combine lineaveraged plasma density measurements from a microwave interferometer plasma diagnostic system with vacuum vessel pressure measurements from ion gauges, with various other measurements, and with preprogrammed reference histories, to determine commands to piezoelectric gas injection valves

  7. TFTR tritium handling concepts

    International Nuclear Information System (INIS)

    Garber, H.J.

    1976-01-01

    The Tokamak Fusion Test Reactor, to be located on the Princeton Forrestal Campus, is expected to operate with 1 to 2.5 MA tritium--deuterium plasmas, with the pulses involving injection of 50 to 150 Ci (5 to 16 mg) of tritium. Attainment of fusion conditions is based on generation of an approximately 1 keV tritium plasma by ohmic heating and conversion to a moderately hot tritium--deuterium ion plasma by injection of a ''preheating'' deuterium neutral beam (40 to 80 keV), followed by injection of a ''reacting'' beam of high energy neutral deuterium (120 to 150 keV). Additionally, compressions accompany the beam injections. Environmental, safety and cost considerations led to the decision to limit the amount of tritium gas on-site to that required for an experiment, maintaining all other tritium in ''solidified'' form. The form of the tritium supply is as uranium tritide, while the spent tritium and other hydrogen isotopes are getter-trapped by zirconium--aluminum alloy. The issues treated include: (1) design concepts for the tritium generator and its purification, dispensing, replenishment, containment, and containment--cleanup systems; (2) features of the spent plasma trapping system, particularly the regenerable absorption cartridges, their integration into the vacuum system, and the handling of non-getterables; (3) tritium permeation through the equipment and the anticipated releases to the environment; (4) overview of the tritium related ventilation systems; and (5) design bases for the facility's tritium clean-up systems

  8. Experimental results from detached plasmas in TFTR

    International Nuclear Information System (INIS)

    Strachan, J.D.; Boody, F.P.; Bush, C.E.

    1986-10-01

    Detached plasmas are formed in TFTR which have the principal property of the boundary to the high temperature plasma core being defined by a radiating layer. This paper documents the properties of TFTR ohmic-detached plasmas with a range of plasma densities at two different plasma currents

  9. Plasma-wall interaction: Recent TFTR results and implications on design and construction of limiters

    International Nuclear Information System (INIS)

    Owens, D.K.; Ulrickson, M.A.

    1987-01-01

    The first wall of the Tokamak Fusion Test Reactor (TFTR) consists of a water cooled toroidal belt limiter, a cooled moveable limiter, and cooled protective plates to shield the vacuum vessel from neutral beam shinethrough. Each of these systems consists of Inconel support plates covered with graphite tiles. In addition, there are Inconel and stainless steel bellows cover plates to protect the bellows and the surface pumping system which provides enhanced pumping in the torus and also serves to protect the bellows. These systems are described and the design requirements, simulations and actual thermal and mechanical loads reviewed. The normal and off-normal operating conditions which were considered in the design of the TFTR components include thermal loading during normal and disruptive plasma operation, eddy-current induced mechanical forces and arcing. The failures which have occurred are generally associated with thermal stress rather than mechanical failure due to disruption induced eddy currents. The models which were developed to design the TFTR hardware appear to have worked well as the performance of these systems has generally been satisfactory at loads approaching design limits. The implications of the TFTR experience for reactor design are discussed

  10. Measurement of TFTR D-T radiation shielding efficiency

    International Nuclear Information System (INIS)

    Kugel, H.W.; Ascione G.; Elwood, S.

    1994-01-01

    High power D-T fusion reactor designs presently exhibit complex geometric and material density configurations. Simulations of the radiation shielding required for safe operation and full compliance with all regulatory requirements must include sufficient margin to accommodate uncertainties in material properties and distributions, uncertainties in the final configurations, and uncertainties in approximations employing the homogenization of complex geometries. Measurements of radiation shielding efficiency performed in a realistic D-T tokamak environment can provide empirical guidance for simulating safe, efficient, and cost effective shielding systems for future high power fusion reactors. In this work, the authors present the results of initial measurements of the TFTR radiation shielding efficiency during high power D-T operations with record neutron yields. The TFTR design objective is to limit the total dose-equivalent at the nearest PPPL property lines from all radiation pathways to 10 mrem per calendar year. Compliance with this design objective over a calendar year requires measurements in the presence of typical site backgrounds of about 80 mrem per year

  11. Tritium decontamination of TFTR carbon tiles employing ultra violet light

    International Nuclear Information System (INIS)

    Shu, W.M.; Ohira, S.; Gentile, C.A.; Oya, Y.; Nakamura, H.; Hayashi, T.; Iwai, Y.; Kawamura, Y.; Konishi, S.; Nishi, M.F.; Young, K.M.

    2001-01-01

    Tritium decontamination on the surface of Tokamak Fusion Test Reactor (TFTR) bumper limiter tiles used during the Deuterium-Deuterium (D-D) phase of TFTR operations was investigated employing an ultra violet light source with a mean wavelength of 172 nm and a maximum radiant intensity of 50 mW/cm 2 . The partial pressures of H 2 , HD, C and CO 2 during the UV exposure were enhanced more than twice, compared to the partial pressures before UV exposure. In comparison, the amount of O 2 decreased during the UV exposure and the production of a small amount of O 3 was observed when the UV light was turned on. Unlike the decontamination method of baking in air or oxygen, the UV exposure removed hydrogen isotopes from the tile to vacuum predominantly in forms of gases of hydrogen isotopes. The tritium surface contamination on the tile in the area exposed to the UV light was reduced after the UV exposure. The results show that the UV light with a wavelength of 172 nm can remove hydrogen isotopes from carbon-based tiles at the very surface

  12. Health physics and radioactive waste considerations for the TFTR

    International Nuclear Information System (INIS)

    Rule, K.; Gilbert, J.; Scott, J.

    1995-01-01

    The Tokamak Fusion Test Reactor (TFTR) began high power fusion operations, with tritium, in November of 1993. The operational health physics program involves maintenance on activated materials and tritium contaminated systems. Survey data and findings are collected on routine and special maintenance situations ranging from work on small volume piping to large volume neutral beam systems. The results of radiological measurements are described in relation to the differentiation of elemental tritium to tritium oxide in worker's breathing zones and the associated general work area. The contamination levels, airborne radioactivity, and oil concentrations are also compared. Measurements for gamma radiation are performed to determine personnel access requirements and for comparison to activation and decay models as a planning tools. TFTR presents many unusual challenges with regard to dismantling, packaging and disposal of its components and ancillary systems. A functional time phased radioactive waste generation schedule was developed to enhance project planning. This project will be the first demonstration of the decommissioning of a tritium fueled fusion test reactor

  13. TFTR grounding scheme and ground-monitor system

    International Nuclear Information System (INIS)

    Viola, M.

    1983-01-01

    The Tokamak Fusion Test Reactor (TFTR) grounding system utilizes a single-point ground. It is located directly under the machine, at the basement floor level, and is tied to the building perimeter ground. Wired to this single-point ground, via individual 500 MCM insulated cables, are: the vacuum vessel; four toroidal field coil cases/inner support structure quadrants; umbrella structure halves; the substructure ring girder; radial beams and columns; and the diagnostic systems. Prior to the first machine operation, a ground-loop removal program was initiated. It required insulation of all hangers and supports (within a 35-foot radius of the center of the machine) of the various piping, conduits, cable trays, and ventilation systems. A special ground-monitor system was designed and installed. It actively monitors each of the individual machine grounds to insure that there are no inadvertent ground loops within the machine structure or its ground and that the machine grounds are intact prior to each pulse. The TFTR grounding system has proven to be a very manageable system and one that is easy to maintain

  14. Shielding calculations for the TFTR neutral beam injectors

    International Nuclear Information System (INIS)

    Santoro, R.T.; Lillie, R.A.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1979-07-01

    Two-dimensional discrete ordinates calculations have been performed to determine the location and thickness of concrete shielding around the Tokamak Fusion Test Reactor (TFTR) neutral beam injectors. Two sets of calculations were performed: one to determine the dose equivalent rate on the roof and walls of the test cell building when no injectors are present, and one to determine the contribution to the dose equivalent rate at these locations from radiation streaming through the injection duct. Shielding the side and rear of the neutral beam injector with 0.305 and 0.61 m of concrete, respectively, and lining the inside of the test cell wall with an additional layer of concrete having a thickness of 0.305 m and a height above the axis of deuteron injection of 3.10 m are sufficient to maintain the biological dose equivalent rate outside the test cell to approx. 1 mrem/DT pulse

  15. Measurements of tritium retention and removal on TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Kamperschroer, J.

    1996-05-01

    Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have afforded an opportunity to measure the retention of tritium in a graphite limiter that is subject to erosion, codeposition and high neutron flux. The tritium was injected by both gas puff and neutral beams. The isotopic mix of hydrogenic recycling was measured spectroscopically and the tritium fraction T/(H+D+T) increased to as high as 75%. Some tritium was pumped out during the experimental run and some removed in a subsequent campaign using various clean-up techniques. While the short term retention of tritium was high, various conditioning techniques were successful in removing ∼ 8,000 Ci and restoring the tritium inventory to a level well below the administrative limit

  16. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    1982-03-01

    There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

  17. Control of TFTR during DT operations

    International Nuclear Information System (INIS)

    Pearson, G.G.; Alling, P.D.; Blanchard, W.; Camp, R.A.; Hawryluk, R.J.; Hosea, J.C.; Nagy, A.

    1995-01-01

    Since beginning routine D-T operations in December, 1993, there have been more than 500 DT plasmas and approximately 600,000 Ci of tritium processed through TFTR culminating in greater than 10 MW of fusion power produced in a single discharge in November, 1994. These performance levels were achieved while maintaining the highest levels of personnel and equipment safety. Prior to D-T operations, a Chain of Command structure and a TFTR Shift Supervisor (TFTRSS) position were developed for centralized control of the facility with all subsystems reporting to this position. A comprehensive surveillance system was incorporated such that the TFTR SS could easily review the operational readiness of subsystems for D-T operations. A TFTR SS Station was constructed to facilitate monitoring and control of TFTR. This station includes a camera system, FAX, a networked personal computer, a computerized tritium monitor and control system and a hardware interlock system. In the transition from D-D to D-T operations, TFTR's procedures were reviewed/revised and a number of additional procedures developed for control of activities at the facility. This paper details the equipment, administrative and organizational configurations used for controlling TFTR during D-T operations

  18. Deuterium-tritium TFTR plasmas in the high poloidal beta regime

    International Nuclear Information System (INIS)

    Sabbagh, S.A.; Mauel, M.E.; Navratil, G.A.

    1995-03-01

    Deuterium-tritium plasmas with enhanced energy confinement and stability have been produced in the high poloidal beta, advanced tokamak regime in TFTR. Confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.46 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasma,s exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Fusion power exceeding 6.7 MW with a fusion power gain Q DT = 0.22 has been produced with reduced alpha particle first orbit loss provided by the increased l i

  19. Special remote tooling developed and utilized to tighten TFTR TF coil casing bolts

    International Nuclear Information System (INIS)

    Burgess, T.W.; Walton, G.R.; Meighan, T.G.; Paul, B.L.

    1993-01-01

    Special tooling has been developed and used to tighten toroidal field (TF) coil casing bolts that have loosened from years of Tokamak Fusion Test Reactor (TFTR) operation. Due to their location, many of the TF casing bolts cannot be directly accessed or viewed; their condition was first discovered during unrelated inspections in 1988. Engineering solutions were, sought until 1992, when a remotely operated wrench concept was successfully demonstrated on a TF coil mockup. The concept was developed into several working tools that have successfully been applied to tighten several thousand TF casing bolts during recent scheduled outages. This effort has improved the integrity and reliability of the TF coil system in preparing for the final experimental phase of the TFTR. This paper discusses the design and application of this tooling

  20. Simulation of α-particle redistribution due to sawteeth on TFTR

    International Nuclear Information System (INIS)

    Yi Zhao; White, R.B.

    1996-01-01

    In recent Deuterium-Tritium experiments on the Tokamak Fusion Test Reactor (TFTR), both the Pellet Charge Exchange (PCX) and the alpha Charge Exchange Recombination Spectroscopy (α-CHERS) diagnostics indicate that sawtooth oscillations can cause significant broadening of the fusion alpha radial density profile. The authors investigate this sawtooth mixing phenomenon by applying a Hamiltonian guiding center approach. A model of time evolution of the Kadomtsev-type sawtooth is constructed. The presence of more than one mode in the nonlinear stage of the sawtooth crash is necessary to cause significant broadening of the alpha density profile. Use of numerical equilibria allows us to perform detailed comparisons with TFTR experimental data. The results are in reasonable agreement with α-CHERS and show a broadening of alpha particles similar to that seen in PCX measurements

  1. Tritium processing and management during D-T experiments on TFTR

    International Nuclear Information System (INIS)

    La Marche, P.H.; Anderson, J.L.; Gentile, C.A.; Hawryluk, R.J.; Hosea, J.; Kalish, M.; Kozub, T.; Murray, H.; Nagy, A.; Raftopoulos, S.

    1994-11-01

    TFTR performance has surpassed many of the previous tokamak records. This has been made possible by the use of tritium as fuel for DT plasma discharges. Stable operations of tritium systems provide for safe, routine DT operation of TFTR. In the preparation for DT operation, in the commissioning of the tritium systems and in the operation of the Nuclear Facility several key lessons have been learned. They include: the facility must take the lead in interpreting the applicable regulations and orders and then seek regulator approval; the use of ultra high vacuum technology in tritium system design and construction simplifies and enhances operations and maintenance; and central facility control under a single supervisory position is crucial to safely orchestrate operational and maintenance activities

  2. Decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Walton, G.R.

    1993-01-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) will complete its experimental lifetime with a series of deuterium-tritium pulses in 1994. As a result, the machine structures will become radioactive, and vacuum components will also be contaminated with tritium. Dose rate levels will range from less than 1 mr/h for external structures to hundreds of mr/h for the vacuum vessel. Hence, decommissioning operations will range from hands on activities to the use of remotely operated equipment. After 21 months of cool down, decontamination and decommissioning (D and D) operations will commence and continue for approximately 15 months. The primary objective is to render the test cell complex re-usable for the next machine, the Tokamak Physics Experiment (TPX). This paper presents an overview of decommissioning TFTR and discusses the D and D objectives

  3. Technique for measuring the losses of alpha particles to the wall in TFTR

    International Nuclear Information System (INIS)

    England, A.C.

    1984-03-01

    It is proposed to measure the losses of alpha particles to the wall in the Tokamak Fusion Test Reactor (TFTR) or any large deuterium-tritium (D-T) burning tokamak by a nuclear technique. For this purpose, a chamber containing a suitable fluid would be mounted near the wall of the tokamak. Alpha particles would enter the chamber through a thin window and cause nuclear reactions in the fluid. The material would then be transported through a tube to a remote, low-background location for measurement of the activity. The most favorable reaction suggested here is 10 B(α,n) 13 N, although 14 N(α,γ) 18 F and others may be possible. The system, the sensitivity, the probe design, and the sources of error are described

  4. Absolute calibration of TFTR helium proportional counters

    International Nuclear Information System (INIS)

    Strachan, J.D.; Diesso, M.; Jassby, D.; Johnson, L.; McCauley, S.; Munsat, T.; Roquemore, A.L.; Loughlin, M.

    1995-06-01

    The TFTR helium proportional counters are located in the central five (5) channels of the TFTR multichannel neutron collimator. These detectors were absolutely calibrated using a 14 MeV neutron generator positioned at the horizontal midplane of the TFTR vacuum vessel. The neutron generator position was scanned in centimeter steps to determine the collimator aperture width to 14 MeV neutrons and the absolute sensitivity of each channel. Neutron profiles were measured for TFTR plasmas with time resolution between 5 msec and 50 msec depending upon count rates. The He detectors were used to measure the burnup of 1 MeV tritons in deuterium plasmas, the transport of tritium in trace tritium experiments, and the residual tritium levels in plasmas following 50:50 DT experiments

  5. A Fast Shutdown Technique for Large Tokamaks

    International Nuclear Information System (INIS)

    Fredrickson, E.; Schmidt, G.L.; Hill, K.; Jardin, S.C.

    1999-01-01

    A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-50 deuterium pellets doped with a small ( 0.0005%) concentration of Krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modeling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR

  6. Radial electric fields for improved tokamak performance

    International Nuclear Information System (INIS)

    Downum, W.B.

    1981-01-01

    The influence of externally-imposed radial electric fields on the fusion energy output, energy multiplication, and alpha-particle ash build-up in a TFTR-sized, fusing tokamak plasma is explored. In an idealized tokamak plasma, an externally-imposed radial electric field leads to plasma rotation, but no charge current flows across the magnetic fields. However, a realistically-low neutral density profile generates a non-zero cross-field conductivity and the species dependence of this conductivity allows the electric field to selectively alter radial particle transport

  7. Current configuration and performance of the TFTR computer system

    International Nuclear Information System (INIS)

    Sauthoff, N.R.; Barnes, D.J.; Daniels, R.; Davis, S.; Reid, A.; Snyder, T.; Oliaro, G.; Stark, W.; Thompson, J.R. Jr.

    1986-01-01

    Developments in the TFTR (Tokamak Fusion Test Reactor) computer support system since its startup phases are described. Early emphasis on tokamak process control have been augmented by improved physics data handling, both on-line and off-line. Data acquisition volume and rate have been increased, and data is transmitted automatically to a new VAX-based off-line data reduction system. The number of interface points has increased dramatically, as has the number of man-machine interfaces. The graphics system performance has been accelerated by the introduction of parallelism, and new features such as shadowing and device independence have been added. To support multicycle operation for neutral beam conditioning and independence, the program control system has been generalized. A status and alarm system, including calculated variables, is in the installation phase. System reliability has been enhanced by both the redesign of weaker components and installation of a system status monitor. Development productivity has been enhanced by the addition of tools

  8. Tokamaks (Second Edition)

    Energy Technology Data Exchange (ETDEWEB)

    Stott, Peter [JET, UK (United Kingdom)

    1998-10-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  9. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    Stott, Peter

    1998-01-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  10. Applying neural networks to control the TFTR neutral beam ion sources

    International Nuclear Information System (INIS)

    Lagin, L.

    1992-01-01

    This paper describes the application of neural networks to the control of the neutral beam long-pulse positive ion source accelerators on the Tokamak Fusion Test Reactor (TFTR) at Princeton University. Neural networks were used to learn how the operators adjust the control setpoints when running these sources. The data sets used to train these networks were derived from a large database containing actual setpoints and power supply waveform calculations for the 1990 run period. The networks learned what the optimum control setpoints should initially be set based uon desired accel voltage and perveance levels. Neural networks were also used to predict the divergence of the ion beam

  11. TFTR bumper limiter and final protective plate engineering, fabrication and assembly

    International Nuclear Information System (INIS)

    Helmich, R.C.; Snook, P.G.; Loesser, G.D.; Reilly, T.B.; Trachsel, C.A.

    1986-01-01

    The inner vacuum vessel wall of the Tokamak Fusion Test Reactor (TFTR) is protected from plasma impingement by a bumper limiter and from neutral beam bombardment by protective plates. Engineering problems and solutions relating to Inconel 718, such as welding, machining in the annealed or age-hardened condition, selection of feeds, speeds and the need for rigid tooling are discussed. Vacuum furnace brazing of the 5/16'' Inconel 600 cooling tubing to the backing plates in both horizontal and vertical sections are shown. A detailed description of the plate and tile fabrication and assembly, with manufacturing and management techniques is outlined in this paper

  12. The effect of toroidal field ripple on confined alphas in TFTR D-T plasmas

    International Nuclear Information System (INIS)

    Duong, H.H.; Medley, S.S.

    1996-05-01

    The Pellet Charge Exchange (PCX) diagnostic on the Tokamak Fusion Test Reactor (TFTR) presently measures trapped alpha distribution functions with very small pitch angle (v parallel /v ∼ 0.05) at the midplane. The measured PCX alpha signal exhibits a depletion region near the outboard region. Results of the alpha energy spectra and radial profile suggest stochastic ripple diffusion is the cause of the depletion. Comparison of the ripple stochastization boundary with Goldston-White-Boozer theory also shows the correct functional dependence on alpha energy and q-profile

  13. Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

    1993-01-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility

  14. The Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Schmidt, J.

    1987-01-01

    The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies

  15. An overview on plasma disruption mitigation and avoidance in tokamak

    International Nuclear Information System (INIS)

    He Kaihui; Pan Chuanhong; Feng Kaiming

    2002-01-01

    Plasma disruption, which seems to be unavoidable in Tokamak operation, occurs very fast and uncontrolled. In order to keep Tokamak plasma from disruption and mitigate the disruption frequency, the research on Tokamak plasma major disruption constitutes one of the main topics in plasma physics. The phenomena and processes of the precursor, thermal quench, current quench, VDE, halo current and runaway electrons generation during plasma disruption are analyzed in detail and systematically based on the data obtained from current Tokamaks such as TFTR, JET, JT-60U and ASDEX-U, etc. The methods to mitigate and avoid disruption in Tokamak are also highlighted schematically. Therefore, it is helpful and instructive for plasma disruption research in next generation large Tokamak such as ITER-FEAT

  16. Physics of high performance deuterium-tritium plasmas in TFTR

    International Nuclear Information System (INIS)

    McGuire, K.M.; Batha, S.

    1996-11-01

    During the past two years, deuterium-tritium (D-T) plasmas in the Tokamak Fusion Test Reactor (TFTR) have been used to study fusion power production, isotope effects associated with tritium fueling, and alpha-particle physics in several operational regimes. The peak fusion power has been increased to 10.7 MW in the supershot mode through the use of increased plasma current and toroidal magnetic field and extensive lithium wall conditioning. The high-internal-inductance (high-I i ) regime in TFTR has been extended in plasma current and has achieved 8.7 MW of fusion power. Studies of the effects of tritium on confinement have now been carried out in ohmic, NBI- and ICRF- heated L-mode and reversed-shear plasmas. In general, there is an enhancement in confinement time in D-T plasmas which is most pronounced in supershot and high-I i discharges, weaker in L-mode plasmas with NBI and ICRF heating and smaller still in ohmic plasmas. In reversed-shear discharges with sufficient deuterium-NBI heating power, internal transport barriers have been observed to form, leading to enhanced confinement. Large decreases in the ion heat conductivity and particle transport are inferred within the transport barrier. It appears that higher heating power is required to trigger the formation of a transport barrier with D-T NBI and the isotope effect on energy confinement is nearly absent in these enhanced reverse-shear plasmas. Many alpha-particle physics issues have been studied in the various operating regimes including confinement of the alpha particles, their redistribution by sawteeth, and their loss due to MHD instabilities with low toroidal mode numbers. In weak-shear plasmas, alpha-particle destabilization of a toroidal Alfven eigenmode has been observed

  17. D-T radiation effects on TFTR diagnostics

    International Nuclear Information System (INIS)

    Ramsey, A.T.

    1994-10-01

    For a 50%-50% deuterium-tritium plasma, the neutron production is 80x higher and the total energy release is 200x higher than the same plasma composed only of deuterium. With this increase in radiation, diagnostics which see only negligible amounts of noise during DD operation may find themselves overwhelmed during DT. The neutrons are not only more numerous, but have 6x as much energy, which causes the calculated 2.4x increase in the gamma flux per neutron near TFTR. We report here the effects of this increased radiation on the TFTR diagnostic set. The most noticeable effects are luminescence and transmission losses in fiber optic signal cables. In addition, a plastic fiber near the torus became unusably opaque after a few DT discharges. Silicon detectors show signs of neutron interactions as well as gamma response, and microchannel electron multipliers show an increased background due to the gamma flux. Bolometers show n and γ heating, and the Thomson scattering intensifier gate spark gap was unreliable until the gas pressure was adjusted. All of these effects were anticipated, and in some cases shielding or compensation techniques were used. Compensation fibers work satisfactorily at these radiation levels, and the rapid fall-off of the radiation as one moves away from the machine makes relocation of fibers and other sensitive components very useful. Conventional shielding designs worked when streaming through signal penetrations was properly dealt with. In coming DT campaigns and the generation of new tokamaks, such problems will be more severe. JET anticipates higher dose levels per shot during DT; TPX has 1000 s pulses and ITER presents a particularly difficult challenge. We shall discuss the implications of our results for diagnostics on these machines

  18. Area Safety Program for the tokamak fusion test reactor (TFTR)

    International Nuclear Information System (INIS)

    Rappe, G.M.

    1984-10-01

    Overall the Area Safety Program has proved to be a very successful operation. There is no doubt that a safety program organized through line management is the best way to involve all personnel. Naturally, when the program was first started, there was some criticism and a certain resistance on the part of a few individuals to fully participate. However, once the program was underway and it could be seen that it was working to everyone's advantage, this reluctance disappeared and a spirit of full cooperation is now enjoyed. It is very important that for this success to continue there must be a two way flow of information, both from the Area Safety Coordinators up through line management, and from senior management, with decisions and answers, back down through the management chain with the utmost dispatch. As with all programs, there is still room for improvement. This program has started a review cycle with a view to streamlining certain areas and possibly increasing its scope in others

  19. Plasma diagnostics on large tokamaks

    International Nuclear Information System (INIS)

    Orlinskij, D.V.; Magyar, G.

    1988-01-01

    The main tasks of the large tokamaks which are under construction (T-15 and Tore Supra) and of those which have already been built (TFTR, JET, JT-60 and DIII-D) together with their design features which are relevant to plasma diagnostics are briefly discussed. The structural features and principal characteristics of the diagnostic systems being developed or already being used on these devices are also examined. The different diagnostic methods are described according to the physical quantities to be measured: electric and magnetic diagnostics, measurements of electron density, electron temperature, the ion components of the plasma, radiation loss measurements, spectroscopy of impurities, edge diagnostics and study of plasma stability. The main parameters of the various diagnostic systems used on the six large tokamaks are summarized in tables. (author). 351 refs, 44 figs, 22 tabs

  20. Operation of the repeating pneumatic injector on TFTR and design of an 8-shot deuterium pellet injector

    International Nuclear Information System (INIS)

    Combs, S.K.; Milora, S.L.; Foust, C.R.

    1985-01-01

    The repeating pneumatic hydrogen pellet injector, which was developed at the Oak Ridge National Laboratory (ORNL), has been installed and operated on the Tokamak Fusion Test Reactor (TFTR). The injector combines high-speed extruder and pneumatic acceleration technologies to propel frozen hydrogen isotope pellets repetitively at high speeds. The pellets are transported to the plasma in an injection line that also serves to minimize the gas loading on the torus; the injection line incorporates a fast shutter valve and two stages of guide tubes with intermediate vacuum pumping stations. A remote, stand-alone control and data acquisition system is used for injector and vacuum system operation. In early pellet fueling experiments on TFTR, the injector has been used to deliver deuterium pellets at speeds ranging from 1.0 to 1.5 km/s into plasma discharges. First, single large (nominal 4-mm-dia) pellets provided high densities in TFTR (1.8 x 10 14 cm -3 on axis); after conversion to smaller (nominal 2.7-mm-dia) pellets, up to five pellets were injected at 0.25-s intervals into a plasma discharge, giving a line-averaged density of 1 x 10 14 cm -3 . Operating characteristics and performance of the injector in initial tests on TFTR are presented

  1. PBX/TFTR pellet program PPPL

    International Nuclear Information System (INIS)

    Schmidt, G.

    1986-01-01

    Goals, current results and plans for pellet injection work for the PBX and TFTR programs are outlined. The present PBX injector is a prototype for ORNL 4 pellet condensing injectors. It has demonstrated that pellet injection on PBX can be used to increase overall density and alter the density profile. Future PBX operation requires reliable operation in deuterium and tritium, multiple pellet capability and ability to vary the size of pellets. These goals will require the construction of a new injector similar to the TFTR DPI system. It has also been demonstrated that pellets can efficiently fuel TFTR, producing a clean, high density plasma. Issues which are still outstanding include isotope exchange effects, use of different pellet sizes, optimization of pellet density perturbations and pellet penetration at high beam power

  2. Overview of TFTR transport studies

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Arunasalam, V.; Beer, M.; Bell, M.; Bell, R.; Biglari, H.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.; Cheng, C.Z.; Chu, T.K.; Cohen, S.A.; Cowley, S.; Efthimion, P.C.; Fredrickson, E.; Furth, H.P.; Goldston, R.J.; Greene, G.; Grek, B.; Grisham, L.R.; Hammett, G.; Hill, K.W.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Janos, A.; Jassby, D.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kieras-Phillips, C.; Kilpatrick, S.J.; Kugel, H.; La Marche, P.H.; LeBlanc, B.; Manos, D.M.; Mansfield, D.K.; Mazzucato, E.; McCarthy, M.P.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Monticello, D.; Motley, R.; Mueller, D.; Nazikian, R.; Owens, D.K.; Park, H.; Park, W.; Paul, S.; Perkins, F.; Ramsey, A.T.; Redi, M.H.; Rewoldt, G.; Roquemore, A.L.; Rutherford, P.H.; Schilling, G.; Schivell, J.; Schmidt, G.L.; Scott, S.D.; Stevens, J.; Stratton, B.C.; Stodiek, W.; Synakowski, E.; Tang, W.; Taylor, G.; Timberlake, J.R.; Towner, H.H.; Ulrickson, M.; von Goeler, S.; Wieland, R.; Williams, M.; Wilson, J.R.; Wong, K.L.; Yamada, M.; Yoshikawa, S.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J.; Bush, C.E.; Fonck, R.J; Roberts, D.; Heidbrink, W.; Kesner, J.; Marmar, E.S.; Snipes, J.; Takase, Y.; Terry, J.; Mauel, M.; Navratil, G.A.; Sabbagh, S.; Nagayama, Y.; Pitcher, S.

    1991-10-01

    A review of TFTR plasma transport studies is presented. Parallel transport and the confinement of suprathermal ions are found to be relatively well described by theory. Cross-field transport of the thermal plasma, however, is anomalous with the momentum diffusivity being comparable to the ion thermal diffusivity and larger than the electron thermal diffusivity in neutral beam heated discharges. Perturbative experiments have studied non-linear dependencies in the transport coefficients and examined the role of possible non-local phenomena. The underlying turbulence has been studied using microwave scattering, beam emission spectroscopy and microwave reflectometry over a much broader range in k perpendicular than previously possible. Results indicate the existence of large-wavelength fluctuations correlated with enhanced transport. MHD instabilities set important operational constraints. However, by modifying the current profile using current ramp-down techniques, it has been possible to extend the operating regime to higher values of both var-epsilon β p and normalized β T . In addition, the interaction of MHD fluctuations with fast ions, of potential relevance to α-particle confinement in D-T plasmas, has been investigated. The installation of carbon-carbon composite tiles and improvements in wall conditioning, in particular the use of Li pellet injection to reduce the carbon recycling, continue to be important in the improvement of plasma performance. 96 refs., 16 figs

  3. TFTR tritium inventory accountability system

    International Nuclear Information System (INIS)

    Saville, C.; Ascione, G.; Elwood, S.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Stencel, J.; Voorhees, D.; Tilson, C.

    1995-01-01

    This paper discusses the program, PPPL (Princeton Plasma Physics Laboratory) Material Control and Accountability Plan, that has been implemented to track US Department of Energy's tritium and all other accountable source material. Specifically, this paper details the methods used to measure tritium in various systems at the Tokamak Fusion Test Reactor; resolve inventory differences; perform inventory by difference inside the Tokamak; process and measure plasma exhaust and other effluent gas streams; process, measure and ship scrap or waste tritium on molecular sieve beds; and detail organizational structure of the Material Control and Accountability group. In addition, this paper describes a Unix-based computerized software system developed at PPPL to account for all tritium movements throughout the facility. 5 refs., 2 figs

  4. Recent D-T results on TFTR

    International Nuclear Information System (INIS)

    Johnson, D.W.; Arunasalam, V.

    1995-10-01

    Routine tritium operation in TFTR has permitted investigations of alpha particle physics in parameter ranges resembling those of a reactor core. ICRF wave physics in a DT plasma and the influence of isotopic mass on supershot confinement have also been studied. Continued progress has been made in optimizing fusion power production in TFTR, using extended machine capability and Li wall conditioning. Performance is currently limited by MHD stability. A new reversed magnetic shear regime is being investigated with reduced core transport and a higher predicted stability limit

  5. TFTR neutral-beam power system

    International Nuclear Information System (INIS)

    Winje, R.A.

    1982-10-01

    The TFTR Neutral Beam Power System (NBPS) consists of the accelerator grid power supply and the auxiliary power supplies required to operate the TFTR 120-keV ion sources. The current configuration of the NBPS including the 11-MVA accelerator grid power supply and the Arc and Filament power supplies isolated for operation at accelerator grid voltages up to 120 kV, is described. The prototype NBPS has been assembled at the Princeton Plasma Physics Laboratory and has been operated. The results of the initial operation and the description and resolution of some of the technical problems encountered during the commissioning tests are presented

  6. Impurity pellet injection experiments at TFTR

    International Nuclear Information System (INIS)

    Marmar, E.S.

    1991-01-01

    Impurity (Li and C) pellet experiments, which began at TFTR in 1989, and are expected to continue at least through 1991, have continued to produce new and significant results. The most significant of these are: (1) improvements in TFTR supershots after wall-conditioning by Li pellet injection; (2) accurate measurements of the pitch angle profiles of the internal magnetic field using the polarization angles of line emission from Li + in the pellet ablation cloud; and (3) initial measurements of pitch angle profiles using the tilt of the LI + emission region of the ablation cloud which is stretched out along the field lines

  7. Radiation shielding for TFTR DT diagnostics

    International Nuclear Information System (INIS)

    Ku, L.P.; Johnson, D.W.; Liew, S.L.

    1994-01-01

    The authors illustrate the designs of radiation shielding for the TFTR DT diagnostics using the ACX and TVTS systems as specific examples. The main emphasis here is on the radiation transport analyses carried out in support of the designs. Initial results from the DT operation indicate that the diagnostics have been functioning as anticipated and the shielding designs are satisfactory. The experience accumulated in the shielding design for the TFTR DT diagnostics should be useful and applicable to future devices, such as TPX and ITER, where many similar diagnostic systems are expected to be used

  8. Plan for decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Walton, G.R.

    1993-01-01

    The Tokamak Fusion Test Reactor (TFTR) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D ampersand D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D ampersand D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D ampersand D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D ampersand D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates

  9. Measurements with vertically viewing charge exchange analyzers during ion cyclotron range of frequencies heating in TFTR

    International Nuclear Information System (INIS)

    Kaita, R.; Hammett, G.W.; Gammel, G.; Goldston, R.J.; Medley, S.S.; Scott, S.D.; Young, K.M.

    1988-01-01

    The utility of charge exchange neutral particle analyzers for studying energetic ion distributions in high-temperature plasmas has been demonstrated in a variety of tokamak experiments. Power deposition profiles have been estimated in the Princeton large torus (PLT) from particle measurements as a function of energy and angle during heating in the ion cyclotron range of frequencies (ICRF) and extensive studies of this heating mode are planned for the upcoming operational period in the tokamak fusion test reactor (TFTR). Unlike the horizontally scanning analyzer on PLT, the TFTR system consists of vertical sightlines intersecting a poloidal cross section of the plasma. A bounce-averaged Fokker--Planck program, which includes a quasilinear operator to calculate ICRF-generated energetic ions, is used to simulate the charge exchange flux expected during fundamental hydrogen heating. These sightlines also cross the trajectory of a diagnostic neutral beam (DNB), and it may be possible to observe the fast ion tail during 3 He minority heating, if the DNB is operated in helium for double charge exchange neutralization

  10. Enhanced carbon influx into TFTR supershots

    International Nuclear Information System (INIS)

    Ramsey, A.T.; Bush, C.E.; Dylla, H.F.; Owens, D.K.; Pitcher, C.S.; Ulrickson, M.

    1990-12-01

    Under some conditions, a very large influx of carbon into TFTR occurs during beam injection into low recycling plasmas (the Supershot regime). These carbon ''blooms'' result in serious degradation of plasma parameters. The sources of this carbon have been identified as hot spots on the TFTR bumper limiter at or near the last closed flux surface. Two separate temperature thresholds have been identified. One, at about 1650 degree C, is consistent with radiation enhanced sublimation. The other, at about 2300 degree C, appears to be thermal sublimation of carbon from the limiter. To account for the increased density caused by the blooms, near unity recycling of the carbon at the limiter by physical sputtering is required; this effect is expected from laboratory measurements, and we believe we are seeing it on TFTR. The sources of the carbon blooms are sites which have either loosely attached fragments of limiter material (caused by damage) or surfaces nearly perpendicular to the magnetic field lines. Such surfaces may have local power depositions two orders of magnitude higher than usual. The TFTR team modified the limiter during the opening of Winter 1989--90. The modifications greatly reduced the number and magnitude of the blooms, so that they are no longer a problem

  11. Enhanced carbon influx into TFTR supershots

    International Nuclear Information System (INIS)

    Ramsey, A.T.; Bush, C.E.; Dylla, H.F.; Owens, D.K.; Pitcher, C.S.; Ulrickson, M.A.

    1991-01-01

    Under some conditions, a very large influx of carbon into TFTR occurs during neutral beam injection into low recycling plasmas (the supershot regime). These carbon ''blooms'' result in serious degradation of plasma parameters. The sources of this carbon have been identified as hot spots on the TFTR bumper limiter at or near the last closed flux surface. Two separate temperature thresholds have been identified. One threshold, at about 1650 deg. C, is consistent with radiation enhanced sublimation (RES). The other, at about 2300 deg. C, appears to be thermal sublimation of carbon from the limiter. The carbon influx can be quantitatively accounted for by taking laboratory values for RES rates, making reasonable assumptions about the extent of the blooming area and assuming unity carbon recycling at the limiter. Such high carbon recycling is expected, and it is shown that, in target plasmas at least, it is observed on TFTR. The sources of the carbon blooms are sites which have either loosely attached fragments of limiter material (caused by damage) or surfaces that are nearly perpendicular to the magnetic field lines. Such surfaces may have local power depositions two orders of magnitude higher than usual. The TFTR team modified the limiter during the opening of winter 1989-1990. The modifications greatly reduced the number and magnitude of the blooms, so that they are no longer a problem. (author). 27 refs, 9 figs

  12. Possible neutral beam requirements for TFTR upgrades

    International Nuclear Information System (INIS)

    Prichard, B.A. Jr.; Little, R.; Post, D.E.; Schmidt, J.A.

    1977-01-01

    A discussion is provided of possible neutral beam requirements and constraints for a TFTR upgrade. The time scale is the early 80s and beams of 250 keV D 0 , probably using 65 ampere negative ion sources, existing power supplies and vacuum enclosures would be required

  13. Tokamak Physics Experiment (TPX) power supply design and development

    International Nuclear Information System (INIS)

    Neumeyer, C.; Bronner, G.; Lu, E.; Ramakrishnan, S.

    1995-01-01

    The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This new feature requires a departure from the traditional tokamak power supply schemes. This paper describes the plan for the adaptation of the PPPL/FTR power system facilities to supply TPX. Five major areas are addressed, namely the AC power system, the TF, PF and Fast Plasma Position Control (FPPC) power supplies, and quench protection for the TF and PF systems. Special emphasis is placed on the development of new power supply and protection schemes

  14. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Perry, E.; Chrzanowski, J.; Rule, K.; Viola, M.; Williams, M.; Strykowsky, R.

    1999-01-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling

  15. Mode particle resonances during near-tangential neutral beam injection in large tokamaks

    International Nuclear Information System (INIS)

    Kaita, R.; White, R.B.; Morris, A.W.; Fredrickson, E.D.; McGuire, K.M.; Medley, S.S.; Scott, S.D.

    1988-01-01

    Coherent magnetohydrodynamic modes have been observed during neutral beam injection in TFTR and JET. Periodic bursts of oscillations were detected with several plasma diagnostics, and Fokker-Planck calculations show that the populations of trapped particles in both tokamaks are sufficient to account for fishbone destabilization. Estimates of mode parameters are in reasonable agreement with the experiments, and they indicate that the fishbone mode may continue to affect the performance of intensely heated tokamaks. 13 refs., 1 fig., 1 tab

  16. ICRF heating and current drive experiments on TFTR

    International Nuclear Information System (INIS)

    Rogers, J.H.; Hosea, J.C.; Phillips, C.K.

    1996-01-01

    Recent experiments in the Ion Cyclotron Range of Frequencies (ICRF) at TFTR have focused on the RF physics relevant to advanced tokamak D-T reactors. Experiments performed either tested confinement in reactor relevant plasmas or tested specific ICRF heating scenarios under consideration for reactors. H-minority heating was used to supply identical heating sources for matched D-T and D only L-mode plasmas to determine the species scaling for energy confinement. Second harmonic tritium heating was performed with only thermal tritium ions in an L-mode target plasma, verifying a possible start-up scenario for the International Thermonuclear Experimental Reactor (ITER). Direct electron heating in Enhanced Reverse Shear (ERS) plasmas has been found to delay the back transition out of the ERS state. D-T mode conversion of the fast magnetosonic wave to an Ion Berstein Wave (IBW) for off-axis heating and current drive has been successfully demonstrated for the first time. Parasitic Li 7 cyclotron damping limited the fraction of the power going to the electrons to less than 30%. Similar parasitic damping by Be 9 could be problematic in ITER. Doppler shifted fundamental resonance heating of beam ions and alpha particles has also been observed

  17. Evaluation of a nonevaporable getter pump for tritium handling in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Singleton, M.F.; Griffith, C.M.

    1978-01-01

    Lawrence Livermore Laboratory has tested and evaluated a commercially available getter pump for use with tritium in the Tokamak Fusion Test Reactor (TFTR). The pump contains Zr(84%)--Al in cartridge form with a concentric heating unit. It performed well in all tests, except for frequent heater failures

  18. Long pulse neutral beam system for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Grisham, L.R.; Bowen, O.N.; Dahlgren, F.; Edwards, J.W.; Kamperschroer, J.; Newman, R.; O'Connor, T.; Ramakrishnan, S.; Rossi, G.; Stevenson, T.; Halle, A. von; Wright, K.E.

    1995-01-01

    The Tokamak Physics Experiment (TPX) is planned as a long-pulse or steady-state machine to serve as a successor to the Tokamak Fusion Test Reactor (TFTR). The neutral beam component of the heating and current drive systems will be provided by a TFTR beamline modified to allow operation for pulse lengths of 1000s. This paper presents a brief overview of the conceptual design which has been carried out to determine the changes to the beamline and power supply components that will be required to extend the pulse length from its present limitation of 1s at full power. The modified system, like the present one, will be capable of injecting about 8MW of power as neutral deuterium. The initial operation will be with a single beamline oriented co-directional to the plasma current, but the TPX system design is capable of accommodating an additional co-directional beamline and a counter-directional beamline. ((orig.))

  19. Collaboration on Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Text Reactor. Final report

    International Nuclear Information System (INIS)

    Intrator, T.

    2000-01-01

    This proposal was peer reviewed and funded as a Collaboration on ''Low Phase Speed Radio Frequency Current Drive Experiments at the Tokamak Fusion Test Reactor''. The original plans we had were to carry out the collaboration proposal by including a post doctoral scientist stationed at PPPL. In response to a 60+% funding cut, all expenses were radically pruned. The post doctoral position was eliminated, and the Principal Investigator (T. Intrator) carried out the brunt of the collaboration. Visits to TFTR enabled T. Intrator to set up access to the TFTR computing network, database, and get familiar with the new antennas that were being installed in TFTR during an up to air. One unfortunate result of the budget squeeze that TFTR felt for its last year of operation was that the experiments that we specifically got funded to perform were not granted run time on TFTR., On the other hand we carried out some modeling of the electric field structure around the four strap direct launch Ion Bernstein Wave (IBW) antenna that was operated on TFTR. This turned out to be a useful exercise and shed some light on the operational characteristics of the IBW antenna and its coupling to the plasma. Because of this turn of events, the project was renamed ''Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Test Reactor''

  20. Thermally excited proton spin-flip laser emission in tokamaks

    International Nuclear Information System (INIS)

    Arunasalam, V.; Greene, G.J.

    1993-07-01

    Based on statistical thermodynamic fluctuation arguments, it is shown here for the first time that thermally excited spin-flip laser emission from the fusion product protons can occur in large tokamak devices that are entering the reactor regime of operation. Existing experimental data from TFTR supports this conjecture, in the sense that these measurements are in complete agreement with the predictions of the quasilinear theory of the spin-flip laser

  1. Note for the Mirnov signal analysis in tokamaks

    International Nuclear Information System (INIS)

    Kikuchi, M.

    1985-05-01

    The relation between Mirnov coil signals and the current perturbation on the rational surface is examined analytically by using the approximate Green's function for the case of large aspect ratio circular tokamaks. Satellite island formation, phase modulation effect due to the poloidal variation of the field line pitch, and the shift effect of the plasma column with respect to the center of the vacuum chamber are examined. The detectability of these effects from Mirnov coil signals is discussed for TFTR

  2. In situ calibration of TFTR neutron detectors

    International Nuclear Information System (INIS)

    Hendel, H.W.; Palladino, R.W.; Barnes, C.W.; Diesso, M.; Felt, J.S.; Jassby, D.L.; Johnson, L.C.; Ku, L.; Liu, Q.P.; Motley, R.W.; Murphy, H.B.; Murphy, J.; Nieschmidt, E.B.; Roberts, J.A.; Saito, T.; Strachan, J.D.; Waszazak, R.J.; Young, K.M.

    1990-01-01

    We report results of the TFTR fission detector calibration performed in December 1988. A NBS-traceable, remotely controlled 252 Cf neutron source was moved toroidally through the TFTR vacuum vessel. Detection efficiencies for two 235 U detectors were measured for 930 locations of the neutron point source in toroidal scans at 16 different major radii and vertical heights. These scans effectively simulated the volume-distributed plasma neutron source and the volume-integrated detection efficiency was found to be insensitive to plasma position. The Campbell mode is useful due to its large overlap with the count rate mode and large dynamic range. The resulting absolute plasma neutron source calibration has an uncertainty of ±13%

  3. Two frequency ICRF operation on TFTR

    International Nuclear Information System (INIS)

    Rogers, J.H.; Majeski, R.; Wilson, J.R.; Hosea, J.C.; Schilling, G.; Stevens, J.; Phillips, C.K.

    1993-01-01

    Modifications have been made recently to allow two of the ICRF antennas (bays L and M) on TFTR to operate at either of two frequencies, 43 MHz or 64 MHz. This was accomplished by lengthening the resonant loops (2Λ at 43 MHz, 3Λ at 64 MHz) and replacing the conventional quarter wave impedance transformers with a tapered impedance design. The other two antennas (bays K and N) will operate at a fixed frequency, 43 MHz. The two frequency operation will allow a combination of 3 He-minority and H-minority heating at near full field on TFTR. The higher frequency, 64 MHz, may also be useful in direct electron heating and current drive experiments at lower toroidal fields. Models of the antenna, resonant loops and impedance matching system are presented

  4. First evidence of collective alpha particle effect on TAE modes in the TFTR D-T experiment

    International Nuclear Information System (INIS)

    Wong, K.L.; Schmidt, G.; Batha, S.H.

    1995-08-01

    The alpha particle effect on the excitation of toroidal Alfven eigenmodes (TAE) was investigated in deuterium-tritium (d-t) plasmas in the Tokamak Fusion Test Reactor (TFTR). RF power was used to position the plasma near the instability threshold, and the alpha particle effect was inferred from the reduction of RF power threshold for TAE instability in d-t plasmas. Initial calculations indicate that the alpha particles contribute 10--30% of the total drive in a d-t plasma with 3 MW of peak fusion power

  5. Measurements of the toroidal plasma rotation velocity in TFTR major-radius compression experiments with auxiliary neutral beam heating

    International Nuclear Information System (INIS)

    Bitter, M.; Wong, K.L.; Scott, S.; Hsuan, H.; Grek, B.; Johnson, D.; Tait, G.

    1990-01-01

    The time history of the central toroidal plasma rotation velocity in Tokamak Fusion Test Reactor (TFTR) experiments [Phys. Rev. Lett. 55, 2587 (1985)] with auxiliary heating by neutral deuterium beam injection and major-radius compression has been measured from the Doppler shift of the emitted Ti XXI Kα line radiation. The experiments were conducted for neutral beam powers in the range 2.1--3.8 MW and line-averaged densities in the range 1.8--3.0x10 19 m -2 . The observed rotation velocity increase during compression is consistent with theoretical estimates

  6. Application of sensitivity analysis to a quantitative assessment of neutron cross-section requirements for the TFTR: an interim report

    International Nuclear Information System (INIS)

    Gerstl, S.A.W.; Dudziak, D.J.; Muir, D.W.

    1975-09-01

    A computational method to determine cross-section requirements quantitatively is described and applied to the Tokamak Fusion Test Reactor (TFTR). In order to provide a rational basis for the priorities assigned to new cross-section measurements or evaluations, this method includes quantitative estimates of the uncertainty of currently available data, the sensitivity of important nuclear design parameters to selected cross sections, and the accuracy desired in predicting nuclear design parameters. Perturbation theory is used to combine estimated cross-section uncertainties with calculated sensitivities to determine the variance of any nuclear design parameter of interest

  7. TFTR control and monitoring system (CICADA)

    International Nuclear Information System (INIS)

    Daniels, R.E.

    1981-01-01

    The TFTR Central Instrumentation, Control and Data Acquisition System (CICADA) is described. This is a computer based system, supporting three types of user interfaces and supporting real time, terminal, and batch operations. Over one hundred graphic display generators will be supported by the system, four array processors will greatly increase the analysis capabilities, and closed circuit television will distribute performance data throughout the facility. Approximately twenty thousand points wll be interfaced to the system

  8. 1987 calibration of the TFTR neutron spectrometers

    International Nuclear Information System (INIS)

    Barnes, C.W.; Strachan, J.D.; Princeton Univ., NJ

    1989-12-01

    The 3 He neutron spectrometer used for measuring ion temperatures and the NE213 proton recoil spectrometer used for triton burnup measurements were absolutely calibrated with DT and DD neutron generators placed inside the TFTR vacuum vessel. The details of the detector response and calibration are presented. Comparisons are made to the neutron source strengths measured from other calibrated systems. 23 refs., 19 figs., 6 tabs

  9. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  10. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1983-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

  11. Application of a two fluid theoretical plasma transport model on current tokamak reactor designs

    International Nuclear Information System (INIS)

    Ibrahim, E.; Fowler, T.K.

    1987-06-01

    In this work, the new theoretical transport models to TIBER II design calculations are described and the results are compared with recent experimental data in large tokamaks (TFTR, JET). Tang's method is extended to a two-fluid model treating ions and electrons separately. This allows for different ion and electron temperatures, as in recent low-density experiments in TFTR, and in the TIBER II design itself. The discussion is divided into two parts: (1) Development of the theoretical transport model and (2) calibration against experiments and application to TIBER II

  12. Recent results and near-term expectations in Tokamak fusion research in the U.S., Europe, and Japan

    International Nuclear Information System (INIS)

    Meade, D.

    1993-01-01

    The development of fusion is often thought about in terms of three different activities: scientific feasibility, engineering feasibility, and economic feasibility. This paper discusses the scientific feasibility of fusion. Reactor temperatures, reactor densities and confinement, particle control, plasma power handling, and self-heating are some of the issues examined. Collaboration and results from research at the Tokamak Fusion Test Reactor (TFTR) at Princeton, the JT-60U in Japan, and JET, the Joint European Torus Tokamak in Oxford are presented

  13. Tritium Removal from JET and TFTR Tiles by a Scanning Laser; TOPICAL

    International Nuclear Information System (INIS)

    C.H. Skinner; N. Bekris; J.P. Coad; C.A. Gentile; M. Glugla

    2002-01-01

    Fast and efficient tritium removal is needed for future D-T machines with carbon plasma-facing components. A novel method for tritium release has been demonstrated on co-deposited layers on tiles retrieved from the Tokamak Fusion Test Reactor (TFTR) and from the Joint European Torus (JET). A scanning continuous wave neodymium laser beam was focused to=100 W/mm2 and scanned at high speed over the co-deposits, heating them to temperatures=2000 C for about 10 ms in either air or argon atmospheres. Fiber optic coupling between the laser and scanner was implemented. Up to 87% of the co-deposited tritium was thermally desorbed from the JET and TFTR samples. This technique appears to be a promising in-situ method for tritium removal in a next-step D-T device as it avoids oxidation, the associated de-conditioning of the plasma-facing surfaces, and the expense of processing large quantities of tritium oxide

  14. Tritium Removal from JET and TFTR Tiles by a Scanning Laser

    International Nuclear Information System (INIS)

    Skinner, C.H.; Bekris, N.; Coad, J.P.; Gentile, C.A.; Glugla, M.

    2002-01-01

    Fast and efficient tritium removal is needed for future D-T machines with carbon plasma-facing components. A novel method for tritium release has been demonstrated on co-deposited layers on tiles retrieved from the Tokamak Fusion Test Reactor (TFTR) and from the Joint European Torus (JET). A scanning continuous wave neodymium laser beam was focused to =100 W/mm2 and scanned at high speed over the co-deposits, heating them to temperatures =2000 C for about 10 ms in either air or argon atmospheres. Fiber optic coupling between the laser and scanner was implemented. Up to 87% of the co-deposited tritium was thermally desorbed from the JET and TFTR samples. This technique appears to be a promising in-situ method for tritium removal in a next-step D-T device as it avoids oxidation, the associated de-conditioning of the plasma-facing surfaces, and the expense of processing large quantities of tritium oxide

  15. Simulations of enhanced reversed shear TFTR discharges with lower hybrid current drive

    International Nuclear Information System (INIS)

    Kesner, J.; Bateman, G.

    1996-01-01

    The BALDUR based BBK code permits predictive simulations of time-dependent tokamak discharges and has the capability to include neutral beam heating, pellet injection, bootstrap currents and lower hybrid current drive. BALDUR contains a theory based multi-regime transport model and previous work has shown excellent agreement with both L-mode and supershot TFTR discharges. These simulations reveal that core transport is dominated by η i and trapped electron modes and the outer region by resistive ballooning. We simulate enhanced reverse shear discharges by beginning with a supershot simulation with a reversed shear profile. Similarly to the TFTR experiments the reversed shear profile is obtained through the programming of the current during startup and the freezing in of these profiles by subsequent heating. At the time of transition into the enhanced confinement regime we turn off the η i and trapped-electron mode transport. We examine the further modification of the plasma current profile that can be obtained with the application of lower hybrid current drive. The results of these simulations will be discussed

  16. Foil deposition alpha collector probe for TFTR's D-T phase

    International Nuclear Information System (INIS)

    Hermann, H.W.; Darrow, D.S.; Timberlake, J.; Zweben, S.J.; Chong, G.P.; Pitcher, C.S.; Macaulay-Newcombe, R.G.

    1995-03-01

    A new foil deposition alpha collector sample probe has been developed for TFTR's D-T phase. D-T fusion produced alpha particles escaping from the plasma are implanted in nickel foils located in a series of collimating ports on the detector. The nickel foils are removed from the tokamak after exposure to one or more plasma discharges and analyzed for helium content. This detector is intended to provide improved alpha particle energy resolution and pitch angle coverage over existing lost alpha detectors, and to provide an absolutely calibrated cross-check with these detectors. The ability to resolve between separate energy components of alpha particle loss is estimated to be ∼ 20%. A full 360 degree of pitch angle coverage is provided for by 8 channels having an acceptance range of ∼ 53 degree per channel. These detectors will be useful in characterizing classical and anomalous alpha losses and any collective alpha instabilities that may be excited during the D-T campaign of TFTR

  17. Analysis of toroidal rotation data for the DIII-D tokamak

    International Nuclear Information System (INIS)

    John, H.St.; Burrell, K.H.; Groebner, R.; DeBoo, J.; Gohil, P.

    1989-01-01

    Both poloidal and toroidal rotation are observed during routine neutral beam heating operation of the DIII-D tokamak. Poloidal rotation results and the empirical techniques used to measure toroidal and poloidal rotation speeds are described by Groebner et al. Here we concentrate on the analysis of recent measurements of toroidal rotation made during diverted, H-mode operation of the DIII-D tokamak during co- and counter-neutral beam injection of hydrogen into deuterium plasmas. Similar studies have been previously reported for Doublet III, ASDEX, TFTR, JET and other tokamaks. (author) 13 refs., 4 figs

  18. Charge-exchange neutral hydrogen measurements in TFTR using Pd-MOS microsensors

    International Nuclear Information System (INIS)

    Bastasz, R.; Kilpatrick, S.J.; Ruzic, D.N.

    1991-06-01

    An array of Pd-metal-oxide semiconductor (Pd-MOS) diodes has been used to monitor the fluence and energy of charge-exchange neutral hydrogen isotopes striking the wall of the Tokamak Fusion Test Reactor (TFTR). The array was positioned 4 cm behind the graphite-tiled wall at the toroidal midplane and exposed to several hundred plasma discharges. Hydrogen isotopes striking the Pd-MOS diodes were detected by measuring the leakage current, which is affected by the presence of these species at the Pd/SiO 2 interface. It was found that the midplane flux strongly increased for neutral-beam heated plasmas and correlated with co-injected neutral beam power. The majority of the neutral flux was <50 eV in energy but its energy distribution extended to above 500 eV. 20 refs., 4 figs

  19. Distributed process control system for remote control and monitoring of the TFTR tritium systems

    International Nuclear Information System (INIS)

    Schobert, G.; Arnold, N.; Bashore, D.; Mika, R.; Oliaro, G.

    1989-01-01

    This paper reviews the progress made in the application of a commercially available distributed process control system to support the requirements established for the Tritium REmote Control And Monitoring System (TRECAMS) of the Tokamak Fusion Test REactor (TFTR). The system that will discussed was purchased from Texas (TI) Instruments Automation Controls Division), previously marketed by Rexnord Automation. It consists of three, fully redundant, distributed process controllers interfaced to over 1800 analog and digital I/O points. The operator consoles located throughout the facility are supported by four Digital Equipment Corporation (DEC) PDP-11/73 computers. The PDP-11/73's and the three process controllers communicate over a fully redundant one megabaud fiber optic network. All system functionality is based on a set of completely integrated databases loaded to the process controllers and the PDP-11/73's. (author). 2 refs.; 2 figs

  20. Analysis of the TFTR toroidal field power supply and its interactions with other loads

    International Nuclear Information System (INIS)

    Newell, W.E.

    1976-01-01

    The rectifiers which supply the four major pulsed loads of the Tokamak Fusion Test Reactor (TFTR) share two flywheel generators. Thus there is a possibility of significant interaction between these rectifiers by way of the notched voltage waveforms which they create at the generator terminals. This paper presents an analysis of the build up of current in the toroidal field (TF) coil, which is the largest load. From this analysis, the notched waveform caused by the TF rectifier is derived and its effect on the other rectifiers is investigated. It is concluded that with the present conceptual design parameters, the external effects of the interactions are likely to be small. However, the internal control circuits of the rectifiers must be carefully designed to minimize those effects

  1. Conceptual thermal-mechanical design of the TFTR first wall armor against neutral beam impingement

    International Nuclear Information System (INIS)

    Chi, J.W.H.; Flaherty, R.

    1976-01-01

    The Tokamak Fusion Test Reactor (TFTR) is designed to operate in a pulsed mode with relatively low duty cycles. Each pulse consists of a short plasma heat-up period, a reaction period, followed by a relatively long cooldown period. Plasma heating is accomplished by ohmic heating by a current induced change in the magnetically linked ohmic heating coils, followed by neutral beam injection for further preheat and the initiation of fusion reactions. During normal operation, the bulk of the neutral beam energy will be absorbed by the plasma, while the remainder will impinge on the vacuum vessel wall. The amount of thermal energy deposited on an unprotected wall is expected to be excessive, limiting the frequency of pulses and requiring frequent wall replacement. A faulted condition would cause penetration of an unprotected wall. As a consequence, a wall armoring (or liner) concept was developed to protect the vacuum vessel wall and to permit ease of liner replacement

  2. Review of recent D-T experiments from TFTR

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, G.; Bateman, G.

    1995-01-01

    An extensive set of deuterium-tritium (D-T) experiments has been carried out on the Tokamak Fusion Test Reactor (TFTR), using nearly equal concentrations of deuterium and tritium. The fusion power has been increased to 9.3 MW, using 34 MW of neutral-beam heating, in a supershot discharge and to 6.7 MW in a high-pp discharge following a current rampdown. Extensive lithium pellet injection has increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-pp discharges. The energy confinement time, τ E , was observed to increase in D-T, relative to D plasmas, by 20% and the n i (0)Ti(0)τ E product by 55%. The improvement in thermal confinement was caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. ICRF heating of a D-T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. The TFIR experiments were able to challenge and confirm several of the underlying assumptions of the ITER design

  3. High speed, locally controlled data acquisition system for TFTR

    International Nuclear Information System (INIS)

    Feng, H.K.; Bradish, G.J.

    1983-01-01

    A high speed, locally controlled, data acquisition and transmission system has been developed by the CICADA (Central Instrumentation Control and Data Acquisition) Group for extracting certain timecritical data during a TFTR pulse and passing it to the control room, 1000 feet distant, to satisfy realtime requirements of frequently sampled variables. The system is designed to utilize any or all of the standard CAMAC (Computer Automated Measurement and Control) modules now employed on the CAMAC links for retrieval of the main body of data, but to operate them in a much faster manner than in a standard CAMAC system. To do this, a pre-programmable ROM sequencer is employed as a controller to transmit commands to the modules at intervals down to one microsecond, replacing the usual CAMAC dedicated computer, and increasing the command rate by an order of magnitude over what could be sent down a Branch Highway. Data coming from any number of channels originating within a single CAMAC ''crate'' is then time-multiplexed and transmitted over a single conductor pair in bi-phase at a 2.5 MHz bit rate using Manchester coding techniques. Benefits gained from this approach include: Reduction in the number of conductors required, elimination of line-to-line skew found in parallel transmission systems, and the capability of being transformer coupled or transmitted over a fiber optic cable to avoid safety hazards and ground loops. The main application for this system so far has been as the feedback path in this closed loop control of currents through the Tokamak's field coils. The paper will treat the system's various applications

  4. TFTR centralized torus interface valve control system

    International Nuclear Information System (INIS)

    Pearson, G.G.; Olsen, D.H.

    1983-01-01

    A system developed especially for the TFTR to monitor and control the interface between the vacuum vessel and associated diagnostics will be described in this paper. Diagnostics which must be connected to the machine vacuum are required to do so through a Torus Interface Valve (TIV). Two types of TIV's are used on TFTR. The first type is a non-latching valve which must be held in the opened position by a sustained OPEN command, returning automatically to the closed position when the OPEN command is removed. This type of TIV is used on all systems which never insert a probe into the vacuum vessel through the TIV. The second type of TIV is a latching valve which requires a momentary OPEN command to open and a momentary CLOSE command to close. Each TIV is linked to its own dedicated logic controller. Each logic controller is hardwired to the appropriate TIV OPEN/CLOSED limit switches, probe IN/OUT limit switches, TFTR vacuum vessel pressure setpoint switches, and diagnostic pressure setpoint switches. The logic controller can be configured for local (push-button) or remote (computer) control. Each controller has a uniquely coded keyswitch to determine the configuration. Whether under local or remote control, all OPEN and CLOSE commands must be approved by the TIV controller (TIVC). In the case of systems with probes, the controller must receive a positive indication that the probe is completely backed out before a CLOSE command will be transmitted from the TIVC to the TIV. Before a valve will be opened by a controller, the differential pressure across the valve must be within certain limits

  5. Alpha particle collective Thomson scattering in TFTR

    International Nuclear Information System (INIS)

    Machuzak, J.S.; Woskov, P.P.; Rhee, D.Y.; Gilmore, J.; Bindslev, H.

    1993-01-01

    A collective Thomson scattering diagnostic is being implemented on TFTR to measure alpha particle, energetic and thermal ion densities and velocity distributions. A 60 GHz, 0.1-1 kW gyrotron will be used as the transmitter source, and the scattering geometry will be perpendicular to the magnetic field in the extraordinary mode polarization. An enhanced scattered signal is anticipated from fluctuations in the lower hybrid frequency range with this scattering geometry. Millimeter wave collective Thomson scattering diagnostics have the advantage of larger scattering angles to decrease the amount of stray light, and long, high power, modulated pulses to obtain improved signal to noise through synchronous detection techniques

  6. TFTR diagnostic control and data acquisition system

    International Nuclear Information System (INIS)

    Sauthoff, N.R.; Daniels, R.E.; PPL Computer Division

    1985-01-01

    General computerized control and data-handling support for TFTR diagnostics is presented within the context of the Central Instrumentation, Control and Data Acquisition (CICADA) System. Procedures, hardware, the interactive man--machine interface, event-driven task scheduling, system-wide arming and data acquisition, and a hierarchical data base of raw data and results are described. Similarities in data structures involved in control, monitoring, and data acquisition afford a simplification of the system functions, based on ''groups'' of devices. Emphases and optimizations appropriate for fusion diagnostic system designs are provided. An off-line data reduction computer system is under development

  7. TPX/TFTR Neutral Beam energy absorbers

    International Nuclear Information System (INIS)

    Dahlgren, F.; Wright, K.; Kamperschroer, J.; Grisham, L.; Lontai, L.; Peters, C.; VonHalle, A.

    1993-01-01

    The present beam energy absorbing surfaces on the TFTR Neutral Beams such as Ion Dumps, Calorimeters, beam defining apertures, and scrapers, are simple water cooled copper plates which wee designed to absorb (via their thermal inertia) the incident beam power for two seconds with a five minute coal down interval between pulses. These components are not capable of absorbing the anticipated beam power loading for 1000 second TPX pulses and will have to be replaced with an actively cooled design. While several actively cooled energy absorbing designs were considered,, the hypervapotron elements currently being used on the JET beamlines were chosen due to their lower cooling water demands and reliable performance on JET

  8. Compact Ignition Tokamak conventional facilities optimization

    International Nuclear Information System (INIS)

    Commander, J.C.; Spang, N.W.

    1987-01-01

    A high-field ignition machine with liquid-nitrogen-cooled copper coils, designated the Compact Ignition Tokamak (CIT), is proposed for the next phase of the United States magnetically confined fusion program. A team of national laboratory, university, and industrial participants completed the conceptual design for the CIT machine, support systems and conventional facilities. Following conceptual design, optimization studies were conducted with the goal of improving machine performance, support systems design, and conventional facilities configuration. This paper deals primarily with the conceptual design configuration of the CIT conventional facilities, the changes that evolved during optimization studies, and the revised changes resulting from functional and operational requirements (F and ORs). The CIT conventional facilities conceptual design is based on two premises: (1) satisfaction of the F and ORs developed in the CIT building and utilities requirements document, and (2) the assumption that the CIT project will be sited at the Princeton Plasma Physics Laboratory (PPPL) in order that maximum utilization can be made of existing Tokamak Fusion Test Reactor (TFTR) buildings and utilities. The optimization studies required reevaluation of the F and ORs and a second look at TFTR buildings and utilities. Some of the high-cost-impact optimization studies are discussed, including the evaluation criteria for a change from the conceptual design baseline configuration. The revised conventional facilities configuration are described and the estimated cost impact is summarized

  9. Comparisons of calculated and measured spectral distributions of neutrons from a 14-MeV neutron source inside the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Barnes, J.M.; Alsmiller, R.G. Jr.; Emmett, M.B.; Drischler, J.D.

    1985-12-01

    A recent paper presented neutron spectral distributions (energy greater than or equal to0.91 MeV) measured at various locations around the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The neutron source for the series of measurements was a small D-T generator placed at various positions in the TFTR vacuum chamber. In the present paper the results of neutron transport calculations are presented and compared with these experimental data. The calculations were carried out using Monte Carlo methods and a very detailed model of the TFTR and the TFTR test cell. The calculated and experimental fluences per unit energy are compared in absolute units and are found to be in substantial agreement for five different combinations of source and detector positions

  10. DT results of TFTR's alpha collector

    International Nuclear Information System (INIS)

    Herrmann, H.W.; Zweben, S.J.; Darrow, D.S.; Timberlake, J.R.; Macaulay-Newcombe, R.G.

    1996-01-01

    An escaping alpha collector probe has been developed for TFTR's DT phase to complement the results of the lost alpha scintillator detectors which have been operating on TFTR since 1988. Measurements of the energy distribution of escaping alphas have been made by measuring the range of alphas implanted into nickel foils located within the alpha collector. Exposed samples have been analyzed for 4 DT plasma discharges at plasma currents of 1.0 and 1.8 MA. The results at 1.0 MA are in good agreement with predictions for first orbit alpha loss at 3.5 MeV. The 1.8 MA results, however, indicate a large anomalous loss of partially thermalized alphas at an energy ∼30% below the birth energy and at a total fluence nearly an order of magnitude above expected first orbit loss. This anomalous loss is not observed with the lost alpha scintillator detectors in DT plasmas but does resemble the anomalous delayed loss seen in DD plasmas. Several potential explanations for this loss process are examined. None of the candidate explanations proposed thus far are fully consistent with the anomalous loss observations

  11. Impurity pellet injection experiments at TFTR

    International Nuclear Information System (INIS)

    Marmar, E.S.

    1992-01-01

    Impurity (Li and C) pellet injection experiments on TFTR have produced a number of new and significant results. (1) We observe reproducible improvements of TFTR supershots after wall-conditioning by Li pellet injection ('lithiumization'). (2) We have made accurate measurements of the pitch angle profiles of the internal magnetic field using two novel techniques. The first measures the internal field pitch from the polarization angles of Li + line emission from the pellet ablation cloud, while the second measures the pitch angle profiles by observing the tilt of the cigar-shaped Li + emission region of the ablation cloud. (3) Extensive measurements of impurity pellet penetration into plasmas with central temperatures ranging from ∼0.3 to ∼7 keV have been made and compared with available theoretical models. Other aspects of pellet cloud physics have been investigated. (4) Using pellets as a well defined perturbation has allowed study of transport phenomena. In the case of small pellet perturbations, the characteristics of the background plasmas are probed, while with large pellets, pellet induced effects are clearly observed. These main results are discussed in more detail in this paper

  12. Simulations of DT experiments in TFTR

    International Nuclear Information System (INIS)

    Budny, R.; Bell, M.G.; Biglari, H.; Bitter, M.; Bush, C.; Cheng, C.Z.; Fredrickson, E.; Grek, B.; Hill, K.W.; Hsuan, H.; Janos, A.; Jassby, D.L.; Johnson, D.; Johnson, L.C.; LeBlanc, B.; McCune, D.C.; Mikkelsen, D.R.; Park, H.; Ramsey, A.T.; Sabbagh, S.A.; Scott, S.; Schivell, J.; Strachan, J.D.; Stratton, B.C.; Synakowski, E.; Taylor, G.; Zarnstorff, M.C.; Zweben, S.J.

    1991-12-01

    A transport code (TRANSP) is used to simulate future deuterium-tritium experiments (DT) in TFTR. The simulations are derived from 14 TFTR DD discharges, and the modeling of one supershot is discussed in detail to indicate the degree of accuracy of the TRANSP modeling. Fusion energy yields and α-particle parameters are calculated, including profiles of the α slowing down time, average energy, and of the Alfven speed and frequency. Two types of simulations are discussed. The main emphasis is on the DT equivalent, where an equal mix of D and T is substituted for the D in the initial target plasma, and for the D O in the neutral-beam injection, but the other measured beam and plasma parameters are unchanged. This simulation does not assume that α heating will enhance the plasma parameters, or that confinement will increase with T. The maximum relative fusion yield calculated for these simulations is Q DT ∼ 0.3, and the maximum α contribution to the central toroidal β is β α (0) ∼ 0.5%. The stability of toroidicity-induced Alfven eigenmodes (TAE) and kinetic ballooning modes (KBM) is discussed. The TAE mode is predicted to become unstable for some of the equivalent simulations, particularly after the termination of neutral beam injection. In the second type of simulation, empirical supershot scaling relations are used to project the performance at the maximum expected beam power. The MHD stability of the simulations is discussed

  13. AC distribution system for TFTR pulsed loads

    International Nuclear Information System (INIS)

    Carroll, R.F.; Ramakrishnan, S.; Lemmon, G.N.; Moo, W.I.

    1977-01-01

    This paper outlines the AC distribution system associated with the Tokamak Fusion Test Reactor and discusses the significant areas related to design, protection, and equipment selection, particularly where there is a departure from normal utility and industrial applications

  14. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    Furth, H.P.

    1986-05-01

    A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime

  15. Orbit effects on impurity transport in a rotating tokamak plasma

    International Nuclear Information System (INIS)

    Wong, K.L.; Cheng, C.Z.

    1988-05-01

    Particle orbits in a rotating tokamak plasma are calculated from the equation of motion in the frame that rotates with the plasma. It is found that heavy particles in a rotating plasma can drift away from magnetic surfaces significantly faster with a higher bounce frequency, resulting in a diffusion coefficient much larger than that for a stationary plasma. Particle orbits near the surface of a rotating tokamak are also analyzed. Orbit effects indicate that more impurities can penetrate into a plasma rotating with counter-beam injection. Particle simulation is carried out with realistic experimental parameters and the results are in qualitative agreement with some experimental observations in the Tokamak Fusion Test Reactor (TFTR). 19 refs., 15 figs

  16. Analysis of IBW experiments on TFTR

    International Nuclear Information System (INIS)

    LeBlanc, B.P.; Bell, R.E.; Bush, C.E.; Cesario, R.; Hanson, G.R.; Hosea, J.; Majeski, R.; Ono, M.; Paoletti, F.; Phillips, C.K.; Rogers, J.H.; Schilling, G.; Wilson, J.R.

    1997-01-01

    A direct launch IBW antenna has been commissioned during the last TFTR experimental campaign. While we did observed IBW-induced poloidal drive, we did not reproduce the CH mode. In this first cut analysis, we concentrate on discharges with hydrogenic resonant species (D or T) combining IBW and neutral beam heating (NBI) at 76 MHz. The experimental data suggest poor power coupling to the main plasma as a limiting factor. A ray tracing code computes the power deposition and results are fed in data reduction code TRANSP to ascertain the coupling efficiency. The density increase observed during IBW is in part caused by influx of impurity, in particular during the latter part of the RF pulse. copyright 1997 American Institute of Physics

  17. Temporary fire sealing of penetrations on TFTR

    International Nuclear Information System (INIS)

    Hondorp, H.L.

    1981-02-01

    The radiation shielding provided for TFTR for D-D and D-T operation will be penetrated by numerous electrical and mechanical services. Eventually, these penetrations will have to be sealed to provide the required fire resistance, tritium sealability, pressure integrity and radiation attenuation. For the initial hydrogen operation, however, fire sealing of the penetrations in the walls and floor is the primary concern. This report provides a discussion of the required and desirable properties of a temporary seal which can be used to seal these penetrations for the hydrogen operation and then subsequently be removed and replaced as required for the D-D and D-T operations. Several candidate designs are discussed and evaluated and recommendations are made for specific applications

  18. Particle reflection and TFTR neutral beam diagnostics

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Grisham, L.R.; Kugel, H.W.; O'Connor, T.E.; Newman, R.A.; Stevenson, T.N.; von Halle, A.; Williams, M.D.

    1992-04-01

    Determination of two critical neutral beam parameters, power and divergence, are affected by the reflection of a fraction of the incident energy from the surface of the measuring calorimeter. On the TFTR Neutral Beam Test Stand, greater than 30% of the incident power directed at the target chamber calorimeter was unaccounted for. Most of this loss is believed due to reflection from the surface of the flat calorimeter, which was struck at a near grazing incidence (12 degrees). Beamline calorimeters, of a ''V''-shape design, while retaining the beam power, also suffer from reflection effects. Reflection, in this latter case, artificially peaks the power toward the apex of the ''V'', complicating the fitting technique, and increasing the power density on axis by 10 to 20%; an effect of import to future beamline designers. Agreement is found between measured and expected divergence values, even with 24% of the incident energy reflected

  19. Visible imaging of edge fluctuations in TFTR

    International Nuclear Information System (INIS)

    Zweben, S.J.; Medley, S.S.

    1989-03-01

    Images of the visible light emission from the inner wall region of TFTR have been made using a rapidly gated, intensified TV camera. Strong ''filamentation'' of the neutral deuterium Dα light is observed when the camera gating time is <100 μsec during neutral-beam-heated discharges. These turbulent filaments vary in position randomly vs. time and have a poloidal wavelength of ∼3-5 cm which is much shorter than their parallel wavelength of ∼100 cm. A second and new type of edge fluctuation phenomenon, which we call a ''merfe,'' is also described. Merfes are a regular poloidal pattern of toroidally symmetric, small-scale marfes which move away from the inner midplane during the current decay after neutral beam injection. Some tentative interpretations of these two phenomena are presented. 27 refs., 8 figs

  20. Soft x-ray tomography on TFTR

    International Nuclear Information System (INIS)

    Kuo-Petravic, G.

    1988-12-01

    The tomographic method used for deriving soft x-ray local emissivities on TFTR, using one horizontal array of 60 soft x-ray detectors, is described. This method, which is based on inversion of Fourier components and subsequent reconstruction, has been applied to the study of a sawtooth crash. A flattening in the soft x-ray profile, which we interpret as an m = 1 island, is clearly visible during the precursor phase and its location and width correlate well with those from electron temperature profiles reconstructed from electron cyclotron emission measurement. The limitations of the Fourier method, due notably to the aperiodic nature of the signals in the fast crash phase and the difficulty of obtaining accurately the higher Fourier harmonics, are discussed. 9 refs., 13 figs

  1. Thermal Response of Tritiated Codeposits from JET and TFTR to Transient Heat Pulses

    International Nuclear Information System (INIS)

    Skinner, C.H.; Bekrisl, N.; Coad, J.P.; Gentile, C.A.; Hassanein, A.; Reiswig, R.; Willms, S.

    2002-01-01

    High heat flux interactions with plasma-facing components have been studied at microscopic scales. The beam from a continuous wave neodymium laser was scanned at high speed over the surface of graphite and carbon fiber composite tiles that had been retrieved from TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) after D-T plasma operations. The tiles have a surface layer of amorphous hydrogenated carbon that was co-deposited during plasma operations, and laser scanning has released more than 80% of the co-deposited tritium. The temperature rise of the co-deposit was much higher than that of the manufactured material and showed an extended time history. The peak temperature varied dramatically (e.g., 1,436 C compared to >2,300 C), indicating strong variations in the thermal conductivity to the substrate. A digital microscope imaged the co-deposit before, during, and after the interaction with the laser and revealed 100-micron scale hot spots during the interaction. Heat pulse durations of order 100 ms resulted in brittle destruction and material loss from the surface, whilst a duration of =10 ms showed minimal changes to the co-deposit. These results show that reliable predictions for the response of deposition areas to off-normal events such as ELMs (edge-localized modes) and disruptions in next-step devices need to be based on experiments with tokamak generated co-deposits

  2. Diffusion of alpha-like MeV ions in TFTR

    International Nuclear Information System (INIS)

    Boivin, R.L.; Zweben, S.J.; Chang, C.S.; Hammett, G.; Mynick, H.E.; White, R.B.

    1991-01-01

    Single particle confinement of alpha particles is of crucial importance in reactor-grade tokamaks like BPX and ITER. Besides the well-known process of first-orbit losses, mechanisms that could lead to significant loss of alpha particles are turbulence-induced diffusion and toroidal field ripple stochastic diffusion. These two mechanisms have been separately studied in TFTR using two different detectors (one at the bottom of the machine and the other near the outer midplane) which can detect escaping charged fusion products, namely the 1 MeV triton and the 3 MeV proton in D-D plasmas (and also the 3.5 MeV alpha in D-T). The main difficulty in this type of experiment lies in the necessity of distinguishing the diffusion process from the always-present first-orbit loss-process. In this paper, we show how these two processes can be distinguished using the pitch-angle discrimination of the detectors. The pitch-angle is defined here as the angle of the particle trajectory with respect to the toroidal direction and so is a measure of the ion magnetic moment, μ. Results obtained at the midplane would be the first reported evidence of TF ripple diffusion in a tokamak. (author) 3 refs., 2 figs

  3. Multiple track Doppler-shift spectroscopy system for TFTR neutral beam injectors

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Kugel, H.W.; Reale, M.A.

    1986-09-01

    A Doppler-shift spectroscopy system has been installed on the TFTR neutral beam injection system to measure species composition during both conditioning and injection pulses. Two intensified vidicon detectors and two spectrometers are utilized in a system capable of resolving data from up to twelve ion sources simultaneously. By imaging the light from six ion sources onto one detector, a cost-effective system has been achieved. Fiber optics are used to locate the diagnostic in an area remote from the hazards of the tokamak test cell allowing continuous access, and eliminating the need for radiation shielding of electronic components. Automatic hardware arming and interactive data analysis allow beam composition to be computed between tokamak shots for use in analyzing plasma heating experiments. Measurements have been made using lines of sight into both the neutralizer and the drift duct. Analysis of the data from the drift duct is both simpler and more accurate since only neutral particles are present in the beam at this location. Comparison of the data taken at these two locations reveals the presence of partially accelerated particles possessing an estimated 1/e half-angle divergence of 15 0 and accounting for up to 30% of the extracted power

  4. Mechanical design of the folded waveguide for PBX-M and TFTR

    International Nuclear Information System (INIS)

    Fogelman, C.H.; Bigelow, T.S.; Yugo, J.J.

    1995-01-01

    The folded waveguide (FWG) antenna is an advanced Cyclotron Range of Frequencies launcher being designed at Oak Ridge National Laboratory in collaboration with Princeton Plasma Physics Laboratory. The FWG offers a drastic increase in radio frequency (RF) power density over typical loop antennas. It also results in internal electric fields of much lower magnitude near the plasma. It is scheduled for installation on either the Tokamak Fusion Test Reactor (TFTR) or the Princeton Beta Experiment-Modified (PBX-M) tokamak in January 1996. The design objective is to provide an FWG that can withstand the thermal loads and disruption scenarios and meet the space constants of both machines. The design is also intended to be prototypical for the International Thermonuclear Experimental Reactor (ITER). The FWG is fully retractable, and maintenance operations can be performed while the vessel remains under vacuum. The FWG can operate in fast-wave mode, or it can be retracted, rotated 90 degrees, and reengaged for the ion-Bernstein wave launch. The polarizing plate completely covers the front of the antenna, except for slots cut at every other gap between with plates of other configurations such as a 0-π dipole plate

  5. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

    A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

  6. The Tokamak Fusion Test Reactor D-T modifications and operations

    International Nuclear Information System (INIS)

    1992-01-01

    This Environmental Assessment (EA) was prepared in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended, in support of the Department of Energy's proposal for the Tokamak Fusion Test Reactor (TFTR) D-T program. The objective of the proposed D-T program is to take the initial step in studying the effects of alpha particle heating and transport in a magnetic fusion device. These studies would enable the successful completion of the original TFTR program objectives, and would support the research and development needs of the Burning Plasma Experiment, BPX (formerly the Compact Ignition Tokamak (CIT)) and International Thermonuclear Experimental Reactor (ITER) in the areas of alpha particle physics, tritium retention, alpha particle diagnostic development, and tritium handling

  7. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    an introduction to diagnostics for tokamaks. The complexity of fusion plasmas is attested to by the discovery of new phenomena and new operational regimes as machine size and power increased and the diagnostic tools improved over the forty years of research on magnetic confinement. The history of those discoveries in the devices which have been built worldwide after the results obtained on the first tokamaks at the Kurchatov Institute had been confirmed is outlined in chapters 11-12. Particular emphasis is naturally given to the results from the larger tokamaks: ASDEX Upgrade, DIII-D, TFTR, JT-60/JT-60U and JET. Chapter 13 is devoted to the International Tokamak Experimental Reactor and prospects beyond ITER. Examples of operational regimes and of often unexpected phenomena are the linear and saturated ohmic confinement modes, confinement degradation when auxiliary heating is applied, the high energy confinement mode, the formation of internal transport barriers in weak or negative central shear discharges, sawtooth relaxations, disruptions, multifaceted asymmetric radiation from the edge, edge localised modes, etc. The relevant observations are described very thoroughly with the support of numerous selected figures and their physical interpretation, a major topic of the book, is carefully discussed on the basis of simplified but convincing mathematical models. With respect to the previous edition (1997), a few additions have been introduced; those concern plasma rotation (section 3.13), internal transport barriers (4.14), the role of radial electric field shear (4.19), turbulence simulations (4.21), impurity transport (4.22) and neoclassical drive of tearing modes (7.3). It is my personal feeling that some of those additions should have been somewhat more elaborated. A few pages have finally been added concerning the TCV, START, MAST, NSTX and ASDEX Upgrade tokamaks. With this book, John Wesson offers the fusion community a very precious and thorough survey of

  8. Confinement studies of neutral beam heated discharges in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, M.; Arunasalam, V.; Bell, J.D.; Stauffer, F.; Bell, M.G.; Bitte, M.; Blanchard, W.R.; Boody, F.; Britz, N.

    1985-11-01

    The TFTR tokamak has reached its original machine design specifications (I/sub p/ = 2.5 MA and B/sub T/ = 5.2T). Recently, the D/sup 0/ neutral beam heating power has been increased to 6.3 MW. By operating at low plasma current (I/sub p/ approx. = 0.8 MA) and low density anti n/sub e/ approx. = 1 x 10/sup 19/m/sup -3/), high ion temperatures (9 +- keV) and rotation speeds (7 x 10/sup 5/ m/s) have been achieved during injection. At the opposite extreme, pellet injection into high current plasmas has been used to increase the line-average density to 8 x 10/sup 19/m/sup -3/ and the central density to 1.6 x 10/sup 20/m/sup -3// This wide range of operating conditions has enabled us to conduct scaling studies of the global energy confinement time in both ohmically and beam heated discharges as well as more detailed transport studies of the profile dependence. In ohmic discharges, the energy confinement time is observed to scale linearly with density only up to anti n/sub e/ approx. 4.5 x 10/sup 19/m/sup -3/ and then to increase more gradually, achieving a maximum value of approx. 0.45 s. In beam heated discharges, the energy confinement time is observed to decrease with beam power and to increase with plasma current. With P/sub b/ = 5.6 MW, anti n/sub e/ = 4.7 x 10/sup 19/m/sup -3/, I/sub p/ = 2.2 MA and B/sub T = 4.7T, the gross energy confinement time is 0.22 s and T/sub i/(0) = 4.8 keV. Despite shallow penetration of D/sup 0/ beams (at the beam energy less than or equal to 80 keV with low species yield), tau/sub E/(a) values are as large as those for H/sup 0/ injection, but central confinement times are substantially greater. This is a consequence of the insensitivity of the temperature and safety factor profile shapes to the heating profile. The radial variation of tau/sub E/ is even more pronounced with D/sup 0/ injection into high density pellet-injected plasmas. 25 refs.

  9. Observation of neoclassical transport in reverse shear plasmas on the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Efthimion, P.C.; Goeler, S. von; Houlberg, W.A.

    2001-01-01

    Perturbative experiments on the Tokamak Fusion Test Reactor (TFTR) have investigated the transport of multiple ion species in reverse shear plasmas. The profile evolution of trace tritium and helium, and intrinsic carbon indicate the formation of core particle transport barriers in ERS plasmas. There is an order of magnitude reduction in the particle diffusivity inside the reverse shear region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)

  10. Observation of neoclassical transport in reverse shear plasmas on the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Efthimion, P.C.; Von Goeler, S.; Houlberg, W.A.

    1999-01-01

    Perturbative experiments on the Tokamak Fusion Test Reactor (TFTR) have investigated the transport of multiple ion species in reverse shear plasmas. The profile evolution of trace tritium and helium, and intrinsic carbon indicate the formation of core particle transport barriers in ERS plasmas. There is an order of magnitude reduction in the particle diffusivity inside the reverse shear region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)

  11. ICRF heating and transport of deuterium-tritium plasmas in TFTR

    International Nuclear Information System (INIS)

    Rogers, J.H.; Schilling, G.; Stevens, J.E.; Taylor, G.; Wilson, J.R.; Bell, M.G.; Budny, R.V.; Bretz, N.L.; Darrow, D.; Fredrickson, E.

    1995-02-01

    This paper describes results of the first experiments utilizing high-power ion cyclotron range of frequency (ICRF) to heat deuterium-tritium (D-T) plasmas in reactor-relevant regimes on the Tokamak Fusion Test Reactor (TFTR). Results from these experiments have demonstrated efficient core, second harmonic, tritium beating of D-T supershot plasmas with tritium concentrations ranging from 6%-40%. Significant direct ion heating on the order of 60% of the input radio frequency (rf) power has been observed. The measured deposition profiles are in good agreement with two-dimensional modeling code predictions. Energy confinement in an rf-heated supershot is at least similar to that without rf, and possibly better in the electron channel. Efficient electron heating via mode conversion of fast waves to ion Bernstein waves (IBW) has been demonstrated in ohmic, deuterium-deuterium and DT-neutral beam injection plasmas with high concentrations of minority 3 He (n 3He /n e = 15% - 30%). By changing the 3 He concentration or the toroidal field strength, the location of the mode-conversion radius was varied. The power deposition profile measured with rf power modulation indicated that up to 70% of the power can be deposited on electrons at an off-axis position. Preliminary results with up to 4 MW coupled into the plasma by 90-degree phased antennas showed directional propagation of the mode-converted IBW. Analysis of heat wave propagation showed no strong inward thermal pinch in off-axis heating of an ohmically-heated target plasma in TFTR

  12. A programmatic framework for the Tokamak Physics Experiment (TPX)

    International Nuclear Information System (INIS)

    Thomassen, K.I.; Goldston, R.J.; Neilson, G.H.

    1993-01-01

    Significant advances have been made in the confinement of reactor-grade plasmas, so that the authors are now preparing for experiments at the open-quotes power breakevenclose quotes level in the JET and TFTR experiments. In ITER the authors will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, the authors must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the US domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX the authors can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. The authors can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current US fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program

  13. Preliminary safety analysis report for the TFTR

    International Nuclear Information System (INIS)

    Lind, K.E.; Levine, J.D.; Howe, H.J.

    A Preliminary Safety Analysis Report has been prepared for the Tokamak Fusion Test Reactor. No accident scenarios have been identified which would result in exposures to on-site personnel or the general public in excess of the guidelines defined for the project by DOE

  14. First-wall and limiter conditioning in TFTR

    International Nuclear Information System (INIS)

    Dylla, H.F.; Blanchard, W.R.; Hawryluk, R.J.

    1984-10-01

    A progress report on the experimental studies of vacuum vessel conditioning during the first year of TFTR operation is presented. A previous paper described the efforts expended to condition the TFTR vessel prior to and during the initial plasma start-up experiments. During the start-up phase, discharge cleaning was performed with the vessel at room temperature. For the second phase of TFTR operations, which was directed towards the optimization of ohmically heated plasmas, the vacuum vessel could be heated to 150 0 C. The internal configuration of the TFTR vessel was more complex during the second phase with the addition of a TiC/C moveable limiter array, Inconel bellows cover plates, and ZrAl getter pumps. A quantitative comparison is given on the effectiveness of vessel bakeout, glow discharge cleaning, and pulse discharge cleaning in terms of the total quantity of removed carbon and oxygen, residual gas base pressures and the resulting plasma impurity levels as measured by visible, uv, and soft x-ray spectroscopy. The initial experience with hydrogen isotope changeover in TFTR is presented including the results of the attempt to hasten the changeover time by using a glow discharge to precondition the vessel with the new isotope

  15. Design study of a fusion-driven tokamak hybrid reactor for fissile fuel production. Final report

    International Nuclear Information System (INIS)

    Rose, R.P.

    1979-05-01

    This study evaluated conceptual approaches for a tokamak fusion-driven fuel producing reactor. The conceptual design of this hybrid reactor was based on using projected state-of-the-art technology for the late 1980s. This reactor would be a demonstration plant and, therefore, first-of-a-kind considerations have been included. The conceptual definitions of two alternatives for the fusion driver were evaluated. A Two-Component Tokamak (TCT) concept, based on the TFTR plasma physics parameters, was compared to a Beam-Driven Thermonuclear (BDTN) concept, based on the USSR T-20 plasma physics parameters

  16. The effect of oxygen on the release of tritium during baking of TFTR D-T tiles

    Energy Technology Data Exchange (ETDEWEB)

    Shu, W.M. E-mail: shu@tpl.tokai.jaeri.go.jp; Gentile, C.A.; Skinner, C.H.; Langish, S.; Nishi, M.F

    2002-11-01

    A series of tests involving 10 h baking under the current ITER design conditions (240 deg. C with 933 Pa O{sub 2}) was performed using a cube of a carbon fiber composite tile that had been used in Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium burning operation. The removal rate of the codeposits was about 3 {mu}m/h near the surface and 0.9 {mu}m/h in the deeper region. Total amount of tritium released from the cube during 10 h baking was 202 MBq, while remaining tritium in the cube after baking was 403 MBq. Thus 10 h baking at 240 deg. C with 933 Pa O{sub 2} removed 1/3 of tritium from the cube. After 10 h baking, the tritium concentration on the cube surface also dropped by about 1/3. In addition, some tritium was released from another cube of the tile during baking at 240 deg. C in pure Ar, and a rapid increase of tritium release was observed when the purging gas was shifted from pure Ar to Ar-1%O{sub 2}. When a whole TFTR tile was baked in air at 350 deg. C for 1 h and then at 500 deg. C for 1 h, the ratios of tritium released were 53 and 47%, respectively. Oxygen reacted with carbon to produce carbon monoxide during baking in air.

  17. The effect of oxygen on the release of tritium during baking of TFTR D-T tiles

    International Nuclear Information System (INIS)

    Shu, W.M.; Gentile, C.A.; Skinner, C.H.; Langish, S.; Nishi, M.F.

    2002-01-01

    A series of tests involving 10 h baking under the current ITER design conditions (240 deg. C with 933 Pa O 2 ) was performed using a cube of a carbon fiber composite tile that had been used in Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium burning operation. The removal rate of the codeposits was about 3 μm/h near the surface and 0.9 μm/h in the deeper region. Total amount of tritium released from the cube during 10 h baking was 202 MBq, while remaining tritium in the cube after baking was 403 MBq. Thus 10 h baking at 240 deg. C with 933 Pa O 2 removed 1/3 of tritium from the cube. After 10 h baking, the tritium concentration on the cube surface also dropped by about 1/3. In addition, some tritium was released from another cube of the tile during baking at 240 deg. C in pure Ar, and a rapid increase of tritium release was observed when the purging gas was shifted from pure Ar to Ar-1%O 2 . When a whole TFTR tile was baked in air at 350 deg. C for 1 h and then at 500 deg. C for 1 h, the ratios of tritium released were 53 and 47%, respectively. Oxygen reacted with carbon to produce carbon monoxide during baking in air

  18. Tritium Decontamination of TFTR D-T Graphite Tiles Employing Ultra Violet Light and a Nd:YAG Laser

    International Nuclear Information System (INIS)

    Gentile, C.A.; Skinner, C.H.; Young, K.M.; Ciebiera, L.

    1999-01-01

    The use of an ultra violet (UV) light source (wavelength = 172 nm) and a Nd:YAG Laser for the decontamination of the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium (D-T) tiles will be investigated at the Princeton Plasma Physics Laboratory (PPPL). The development of this form of tritium decontamination may be useful for future D-T burning fusion devices which employ carbon plasma-facing components on the first wall. Carbon tiles retain hydrogen isotopes, and the in-situ tritium decontamination of carbon can be extremely important in maintaining resident in-vessel tritium inventory to a minimum. A test chamber has been designed and fabricated at PPPL. The chamber has the ability to be maintained under vacuum, be baked to 200 *C, and provides sample ports for gas analyses. Tiles from TFTR that have been exposed to D-T plasmas will be placed within the chamber and exposed to either an UV light source or the ND:YAG Laser. The experiment will determine the effectiveness of these two techniques for the removal of tritium. In addition, exposure rates and scan times for the UV light source and/or Nd:YAG Laser will be determined for tritium removal optimization from D-T tiles

  19. Development of the TFTR neutral beam injection system

    International Nuclear Information System (INIS)

    Prichard, B.A. Jr.

    1978-01-01

    The TFTR Neutral Beam Lines are designed to inject 20 MW of 120 keV neutral deuterium atoms into the plasma. This is accomplished using 12 sources, 65 amperes each, mounted in 4 beam lines. The 120 kV sources are being developed by LBL and a prototype beam line which will be tested at Berkeley is being developed as a cooperative effort by LLL and LBL. The implementation of these beam lines has required the development of several associated pieces of hardware. The control and monitoring of the 12 sources will be done via the TFTR computer control system (CICADA) as will other parts of the machine, and software is being developed to condition and operate the sources automatically. The prototype beam line is scheduled to begin operation in the fall of 1978 and all four production beam lines on TFTR in 1982

  20. Discharge cleaning on TFTR after boronization

    International Nuclear Information System (INIS)

    Mueller, D.; Dylla, H.F.; LaMarche, P.H.; Bell, M.G.; Blanchard, W.; Bush, C.E.; Gentile, C.; Hawryluk, R.J.; HIll, K.W.; Janos, A.C.; Jobes, F.C; Owens, D.K.; Pearson, G.; Schivell, J.; Ulrickson, M.A.; Vannoy, C.; Wong, K.L.

    1991-05-01

    At the beginning of the 1990 TFTR experimental run, after replacement of POCO-AXF-5Q graphite tiles on the midplane of the bumper limiter by carbon fiber composite (CFC) tiles and prior to any Pulse Discharge Cleaning (PDC), boronization was performed. Boronization is the deposition of a layer of boron and carbon on the vacuum vessel inner surface by a glow discharge in a diborane, methane and helium mixture. The amount of discharge cleaning required after boronization was substantially reduced compared to that which was needed after previous openings when boronization was not done. Previously, after a major shutdown, about 10 5 low current (∼20 kA) Taylor Discharge Cleaning (TDC) pulses were required before high current (∼400 kA) aggressive Pulse Discharge Cleaning (PDC) pulses could be performed successfully. Aggressive PDC is used to heat the limiters from the vessel bakeout temperature of 150 degrees C to 250 degrees C for a period of several hours. Heating the limiters is important to increase the rate at which water is removed from the carbon limiter tiles. After boronization, the number of required TDC pulses was reduced to <5000. The number of aggressive PDC pulses required was approximately unchanged. 14 refs., 1 tab

  1. Investigation of global Alfven instabilities in TFTR

    International Nuclear Information System (INIS)

    Wong, K.L.; Paul, S.F.; Fredrickson, E.D.; Nazikian, R.; Park, H.K.; Bell, M.; Bretz, N.L.; Budny, R.; Cheng, C.Z.; Cohen, S.; Hammett, G.W.; Jobes, F.C.; Johnson, L.; Meade, D.M.; Medley, S.S.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Synakowski, E.J.; Roberts, D.R.; Sabbagh, S.

    1992-01-01

    Toroidal Alfven Eigenmodes (TAE) were excited by the energetic neutral beam ions tangentially injected into TFTR plasmas at low magnetic field such that the injection velocities were comparable to the Alfven speed. The modes were identified by measurements from Mirnov coils and beam emission spectroscopy (BES). TAE modes appear in bursts whose repetition rate increases with beam power. The neutron emission rate exhibits sawtooth-like behavior and the crashes always coincide with TAE bursts. This indicates ejection of fast ions from the plasma until these modes are stabilized. The dynamics of growth and stabilization was investigated at various plasma current and magnetic field. The results indicate that the instability can effectively clamp the number of energetic ions in the plasma. The observed instability threshold is discussed in the light of recent theories. In addition to these TAE modes, intermittent oscillations at three times the fundamental TAE frequency were observed by Mirnov coils, but no corresponding signal was found in BES. It appears that these high frequency oscillations do not have direct effect on the plasma neutron source strength

  2. Resistive MHD studies of TFTR discharges

    International Nuclear Information System (INIS)

    Hughes, M.H.; Phillips, M.W.; Sabbagh, S.A.; Budny, R.V.

    1991-01-01

    MHD instabilities, thought to be resistive in character, are frequently observed in the supershot operating regime of TFTR (var-epsilon β p ≤ 0.7). These instabilities are always accompanied by substantial degradation of the confinement. Similarly of interest are recent experiments at much larger β p (var-epsilon β p ≤ 1.6), achieved through ramping the current during the beam heating phase of the discharge. In this latter regime the confinement can exceed three times the corresponding L-mode value and the β value normalized to I/aB can be as large as 4.7. Representative discharges from each of these operating regimes have been analyzed using a linear resistive MHD stability code with equilibrium pressure and q profiles obtained initially from the TRANSP analysis code. The main difference between the two types of discharge, as far as stability is concerned is shown to be the shape of the current density profile. The sensitivity to the assumed parameters is discussed. 1 ref

  3. TFTR movable limiter instrumentation and controls

    International Nuclear Information System (INIS)

    Frankenberg, J.; Collins, D.; Kaufmann, D.; Mamoun, A.

    1983-01-01

    The TFTR movable limiter is a single poloidal limiter located within one 18 /SUP o/ segment of the vacuum vessel. It consists of three (3) interconnected inconel backing plates covered with titanium carbide coated graphite tiles. The backing plates are positioned by three independent screw drive actuators. Cooling water is fed through the horizontal port cover to tubes brazed onto the backs of the backing plates. Thermocouples monitor the limiter temperature. (1) and more fully described in refs. (1) and (2). The positioning actuators are driven by independently controlled DC servo motors, controlled either locally or from CICADA. Drive motor shaft position is monitored by chain driven encoders and potentiometers. Limiter blade position can be varied to suit any plasma within the operating range. CICADA is programmed to keep the limiter stroke within safe operating limits. A microprocessor duplicates the CICADA protective function allowing limiter operation without CICADA. The potentiometer signal is sent to an analog computer, which safeguards the limiter against failure of the encoders or the micro-processor. Cooling water flows through the limiter in 3 separate paths, one for each blade. The flow rate and temperature rise through each loop are measured accurately to allow CICADA to calculate the heat into each blade. The water system is also interlocked and alarmed to prevent dumping of water into the vacuum vessel

  4. A tritium vessel cleanup experiment in TFTR

    International Nuclear Information System (INIS)

    Caorlin, M.; Kamperschroer, J.; Owens, D.K.; Voorhees, D.; Mueller, D.; Ramsey, A.T.; La Marche, P.H.; Loughlin, M.J.

    1995-03-01

    A simple tritium cleanup experiment was carried out in TFTR following the initial high power deuterium-tritium discharges in December 1993. A series of 34 ohmic and deuterium neutral beam fueled shots was used to study the removal of tritium implanted into the wall and limiters. A very large plasma was created in each discharge to ''scrub'' an area as large as possible. Beam-fueled shots at 2.5 to 7.5 MW of injected power were used to monitor tritium concentration levels in the plasma by detection of DT-neutrons. The neutron signal decreased by a factor of 4 during the experiment, remaining well above the expected T-burnup level. The amount of tritium recovered at the end of the cleanup was about 8% of the amount previously injected with high power DT discharges. The experience gained suggests that measurements of tritium inventory in the torus are very difficult to execute and require dedicated systems with overall accuracy of 1%

  5. Plasma-material interactions in TFTR

    International Nuclear Information System (INIS)

    Dylla, H.F.; Bell, M.G.; Blanchard, W.R.; Boody, F.P.; Bretz, N.; Budny, R.; Bush, C.E.; Cecchi, J.L.; Cohen, S.A.; Combs, S.K.; Davis, S.L.; Doyle, B.L.; Efthimion, P.C.; England, A.C.; Eubank, H.P.; Fonck, R.; Fredrickson, E.; Grisham, L.R.; Goldston, R.J.; Grek, B.; Groebner, R.; Hawryluk, R.J.; Heifetz, D.; Hendel, H.; Hill, K.W.; Hiroe, S.; Hulse, R.; Johnson, D.; Johnson, L.C.; Kilpatrick, S.; Lamarche, P.H.; Little, R.; Manos, D.M.; Mansfield, D.; Meade, D.M.; Medley, S.S.; Milora, S.L.; Mikkelsen, D.R.; Mueller, D.; Murakami, M.; Nieschmidt, E.; Owens, D.K.; Park, H.; Pontau, A.; Prichard, B.; Ramsey, A.T.; Redi, M.H.; Schivell, J.; Schmidt, G.L.; Scott, S.D.; Sesnic, S.; Shimada, M.; Simpkins, J.E.; Sinnis, J.; Stauffer, F.; Stratton, B.; Tait, G.D.; Taylor, G.; Ulrickson, M.; Von Goeler, S.; Wampler, W.R.; Wilson, K.; Williams, M.; Wong, K.L.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.

    1987-01-01

    This paper presents a summary of plasma-material interactions which influence the operation of TFTR with high current (≤ 2.2 MA) ohmically heated, and high-power (≅ 10 MW) neutral-beam heated plasmas. The conditioning procedures which are applied routinely to the first-wall hardware are reviewed. Fueling characteristics during gas, pellet, and neutral-beam fueling are described. Recycling coefficients near unity are observed for most gas fueled discharges. Gas fueled discharges after helium discharge conditioning of the toroidal bumper limiter, and discharges fueled by neutral beams and pellets, show R e = 5-6x10 19 m -3 ) values of Z eff are ≤ 1.5. Increases in Z eff of ≤ 1 have been observed with neutral beam heating of 10 MW. The primary low Z impurity is carbon with concentrations decreasing from ≅ 10% to e . Oxygen densities tend to increase with n e , and at the ohmic plasma density limit oxygen and carbon concentrations are comparable. Chromium getter experiments and He 2+ /D + plasma comparisons indicate that the limiter is the primary source of carbon and that the vessel wall is a significant source of the oxygen impurity. Metallic impurities, consisting of the vacuum vessel metals (Ni, Fe, Cr) have significant (≅ 10 -4 n e ) concentrations only at low plasma densities (n e 19 m -3 ). The primary source of metallic impurities is most likely ion sputtering from metals deposited on the carbon limiter surface. (orig.)

  6. ICRF stabilization of sawteeth on TFTR

    International Nuclear Information System (INIS)

    Phillips, C.K.; Hosea, J.; Stevens, J.; Wilson, J.R.; Bell, M.; Bitter, M.; Cheng, C.Z.; Darrow, D.; Fredrickson, E.; Hammett, G.W.; Hill, K.; Hsuan, H.; Jassby, D.; McCune, D.; McGuire, K.; Owens, D.K.; Park, H.; Ramsey, A.; Schilling, G.; Schivell, J.; Stratton, B.; Synakowski, E.; Taylor, G.; Towner, H.; White, R.; Zweben, S.; Phillips, M.W.; Hughes, M.; Bush, C.; Goldfinger, R.; Hoffman, D.; Houlberg, W.; Nagayama, Y.; Smithe, D.N.

    1992-01-01

    Results obtained from experiments utilizing high power ICRF (ion cyclotron range of frequency) heating to stabilize sawtooth oscillations on TFTR are reviewed. The key observations include existence of a minimum ICRF power required to achieve stabilization, a dependence of the stabilization threshold on the relative size of the ICRF power deposition profile to the q=1 volume, and a peaking of the equilibrium pressure and current profiles during sawtooth-free phases of the discharges. In addition, preliminary measurements of the poloidal magnetic field profile indicate that q on axis decreases to a value of 0.55±0.15 after a sawtooth-stabilized period of ∼0.5 sec has transpired. The results are discussed in the context of theory, which suggests that the fast ions produced by the ICRF heating suppress sawteeth by stabilizing the m=1 MHD instabilities believed to be the trigger for the sawtooth oscillations. Though qualitative agreement is found between the observations and the theory, further refinement of the theory coupled with more accurate measurements of experimental profiles will be required in order to complete quantitative comparisons

  7. The ICRH tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Perkins, F.W.

    1976-01-01

    A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

  8. Measurements of the toroidal plasma rotation velocity in TFTR major-radius compression experiments with auxiliary neutral beam heating

    International Nuclear Information System (INIS)

    Bitter, M.; Scott, S.; Wong, K.L.

    1986-07-01

    The time history of the central toroidal plasma rotation velocity in Tokamak Fusion Test Reactor (TFTR) experiments with auxiliary heating by neutral deuterium beam injection and major-radius compression has been measured from the Doppler shift of the emitted TiXXI-Kα line radiation. The experiments were conducted for neutral beam powers in the range from 2.1 to 3.8 MW and line-averaged densities in the range from 1.8 to 3.0 x 10 19 m -2 . The observed rotation velocity increase during compression is in agreement with results from modeling calculations which assume classical slowing-down of the injected fast deuterium ions and momentum damping at the rate established in the precompression plasma

  9. Carbon deposition and hydrogen retention in tokamak

    International Nuclear Information System (INIS)

    Tanabe, Tetsuo

    2006-01-01

    The results of measurements on co-deposition of hydrogen isotopes and wall materials, hydrogen retention, redeposition of carbon and deposition of hydrogen on PMI of JT-60U are described. From above results, selection of plasma facing material and ability of carbon wall is discussed. Selection of plasma facing materials in fusion reactor, characteristics of carbon materials as the plasma facing materials, erosion, transport and deposition of carbon impurity, deposition of tritium in JET, results of PMI in JT-60, application of carbon materials to PFM of ITER, and future problems are stated. Tritium co-deposition in ITER, erosion and transport of carbon in tokamak, distribution of tritium deposition on graphite tile used as bumper limiter of TFTR, and measurement results of deposition of tritium on the Mark-IIA divertor tile and comparison between them are described. (S.Y.)

  10. Thermal consequences of plasma disruptions in TFTR and ETF

    International Nuclear Information System (INIS)

    Budny, R.; Ludescher, C.

    1981-01-01

    We studied thermal responses of first walls for TFTR and ETF during plasma disruptions. To model the flux, we assumed the entire kinetic energy is deposited by axisymmetric horizontal displacement of the plasma. The deposition time is a free parameter. In TFTR, the minimum deposition time which does not cause the toroidal limiter to melt is 7 or 14 ms depending on whether or not the limiter is actively cooled. In ETF, the minimum time which does not cause surface melting of the cooling tubes is 80 ms. (author)

  11. Tokamak COMPASS

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan; Křenek, Petr

    2011-01-01

    Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

  12. Alignment of the TFTR bumper limiter

    International Nuclear Information System (INIS)

    Barnes, G.W.; Owens, D.K.; Loesser, G.D.; Ulrickson, M.

    1989-01-01

    The TFTR Bumper Limiter (BL) is an axisymmetric toroidal limiter mounted on the inner wall of the vacuum vessel. It subtends 120 degree poloidally and has a surface area of 22 m 2 . The plasma facing surface consists of 1,000 kg of graphite tiles mounted on watercooled Inconel backing plates. During the initial installation in the Spring of 1985, the limiter surface was aligned to the toroidal magnetic field by mechanical and magnetic measurements to an estimated accuracy of ±2 mm. During subsequent operation, especially in the 1988 run period in which 30 MW of Neutral Beam Injection routinely occurred, several tiles at points on the limiter which protruded slightly into the plasma were severely damaged. The damage, cracked and spalled tiles, is believed to be initiated by high energy disruptions and aggravated by normal high power operation. The damage pattern and temperature rise during normal operation are consistent with this interpretation. A vacuum vessel opening to replace the damaged tiles and realign the limiter was required. The bumper limiter was reshaped to be circular to ±0.5 mm at the midplane by means of mechanical measurements in order to better distribute the heat loads and eliminate hot spots. The ±0.5 mm accuracy is determined by the variation in individual tile thickness which is ±0.5 mm. This paper describes the methods used to mechanically align the limiter and presents evidence based on machine operation with plasma that the limiter is reasonably well aligned with the toroidal field. Future work dealing with the alignment of the total limiter to the toroidal field using mechanical and magnetic measurements and the replacement of a subset of the carbon tiles with carbon-carbon composite material is also discussed. 7 refs., 4 figs

  13. Tritium pellet injector for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Gouge, M.J.; Baylor, L.R.; Combs, S.K.; Fisher, P.W.; Foust, C.R.; Milora, S.L.

    1992-01-01

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  14. Tokamak physics studies using x-ray diagnostic methods

    International Nuclear Information System (INIS)

    Hill, K.W.; Bitter, M.; von Goeler, S.

    1987-03-01

    X-ray diagnostic measurements have been used in a number of experiments to improve our understanding of important tokamak physics issues. The impurity content in TFTR plasmas, its sources and control have been clarified through soft x-ray pulse-height analysis (PHA) measurements. The dependence of intrinsic impurity concentrations and Z/sub eff/ on electron density, plasma current, limiter material and conditioning, and neutral-beam power have shown that the limiter is an important source of metal impurities. Neoclassical-like impurity peaking following hydrogen pellet injection into Alcator C and a strong effect of impurities on sawtooth behavior were demonstrated by x-ray imaging (XIS) measurements. Rapid inward motion of impurities and continuation of m = 1 activity following an internal disruption were demonstrated with XIS measurements on PLT using injected aluminum to enhance the signals. Ion temperatures up to 12 keV and a toroidal plasma rotation velocity up to 6 x 10 5 m/s have been measured by an x-ray crystal spectrometer (XCS) with up to 13 MW of 85-keV neutral-beam injection in TFTR. Precise wavelengths and relative intensities of x-ray lines in several helium-like ions and neon-like ions of silver have been measured in TFTR and PLT by the XCS. The data help to identify the important excitation processes predicted in atomic physics. Wavelengths of n = 3 to 2 silver lines of interest for x-ray lasers were measured, and precise instrument calibration techniques were developed. Electron thermal conductivity and sawtooth dynamics have been studied through XIS measurements on TFTR of heat-pulse propagation and compound sawteeth. A non-Maxwellian electron distribution function has been measured, and evidence of the Parail-Pogutse instability identified by hard x-ray PHA measurements on PLT during lower-hybrid current-drive experiments

  15. Acceleration of beam ions during major radius compression in TFTR

    International Nuclear Information System (INIS)

    Wong, K.L.; Bitter, M.; Hammett, G.W.

    1985-09-01

    Tangentially co-injected deuterium beam ions were accelerated from 82 keV up to 150 keV during a major radius compression experiment in TFTR. The ion energy spectra and the variation in fusion yield were in good agreement with Fokker-Planck code simulations. In addition, the plasma rotation velocity was observed to rise during compression

  16. Mechanical design of epithermal neutron diagnostic for TFTR

    International Nuclear Information System (INIS)

    Groo, R.C.

    1981-01-01

    The mechanical design of the Epithermal Neutron Diagnostic for TFTR is described. This fission detector system measures the time resolution of the neutron flux for folding into the Neutron Activation system and also provides continuous, wide range coverage of all expected fusion reaction rates

  17. Mechanical engineering problems in the TFTR neutral beam system

    International Nuclear Information System (INIS)

    Cannon, D.D.; Bryant, E.H.; Johnson, R.L.; Kim, J.; Queen, C.C.; Schilling, G.

    1975-01-01

    A conceptual design of a prototype beam line for the TFTR Neutral Beam System has been developed. The basic components have been defined, cost estimates prepared, and the necessary development programs identified. Four major mechanical engineering problems, potential solutions and the required development programs are discussed

  18. Measurements of TFTR D-T radiation shielding efficiency

    International Nuclear Information System (INIS)

    Kugel, H.W.; Ascione, G.; Elwood, S.; Gilbert, J.; Ku, L.P.; Levine, J.; Rule, K.; Azziz, N.; Goldhagen, P.; Hajnal, F.

    1994-11-01

    Measurements of neutron and gamma dose-equivalents were performed in the Test Cell, at the outer Test Cell wall, in nearby work areas, and out to the nearest property lines at a distance of 180 m. Argon ionization chambers, moderated 3 He proportional counters, and fission chamber detectors were used to obtain measurements of neutron and gamma dose-equivalents per D-T neutron during individual TFTR discharges. These measured neutron and gamma D-T dose-equivalents per TFTR neutron characterize the effects of local variations in material density resulting from the complex asymmetric site geometry. The measured dose-equivalents per TFTR D-T neutron and the cumulative neutron production were used to determine that the planned annual TFTR neutron production of 1 x 10 21 D-T neutrons is consistent with the design objective of limiting the total dose-equivalent at the property line, from all radiation sources and pathways, to less than 10 mrem per year

  19. Utilization of fusion neutrons in the tokamak fusion test reactor for blanket performance testing and other nuclear engineering experiments

    International Nuclear Information System (INIS)

    Caldwell, C.S.; Pettus, W.G.; Schmotzer, J.K.; Welfare, F.; Womack, R.

    1979-01-01

    In addition to developing a set of reacting-plasma/blanket-neutronics benchmark data, the TFTR fusion application experiments would provide operational experience with fast-neutron dosimetry and the remote handling of blanket modules in a tokamak reactor environment; neutron streaming and hot-spot information invaluable for the optimal design of penetrations in future fusion reactors; and the identification of the most damage-resistant insulators for a variety of fusion-reactor components

  20. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of DandD (Decontamination and Decommissioning) activity

  1. Structural analysis of TFTR TF coils and support structure for 6 Tesla operation

    International Nuclear Information System (INIS)

    Zatz, I.J.; Cargulia, G.; Lontai, L.

    1995-01-01

    The Tokamak Fusion Test Reactor (TFTR), which has been on line since December 1982, has successfully operated at its design Toroidal Field (TF) of 5.2 Tesla. Analysis of test data has indicated that the measured peak D-D neutron power in supershots may be scaled to the fourth power of TF field. Increasing the TF field to 6 Tesla provides the opportunity to explore the possibility of improving the D-T fusion yield, with the use of tritium. This increase in TF field from 5.2 to 6.0 Tesla increases the centering force by 33% and the out-of-plane force by 15% over previous peak operating levels. To examine the impact of the increase in loads on the TF coil, case and supporting structure, finite element analyses (FEA) were performed with and without the presence of loose bolts in the TF case. Note that the loose bolts comprise a fraction of the total number of bolts fastening the TF case sidewalls to the inner and outer rings of the case. Extensive analysis was performed using the FEA results in conjunction with supplementary calculations. Results are presented for the TF case, bolts, copper conductors, insulation, and supporting structure which indicate that the TF coils can successfully operate at 6 Tesla for a reasonable number of pulses

  2. Overcurrent protection for the TFTR neutral beam sources during spark down

    International Nuclear Information System (INIS)

    Praeg, W.F.

    1979-01-01

    The accelerating grid of a neutral beam source (NBS) of the Tokamak Fusion Test Reactor (TFTR) operates at 120 kV and 65 A. The capacitance to ground between the switch tube (ST) and the NBS is C 1 approx. 5 nF (approx. 36 J). The arc and filament power supplies for the NBS float at 120 kV and have a capacitance to ground of C 2 approx. 2 nF (approx. 14 J). When the NBS sparks to ground, C 2 begins to discharge immediately. The ST impedance limits the fault current from the high voltage (HV) power supply to approx. 100 A until it disconnects the power source 1 begins to discharge. During spark down, fault currents are limited with a saturated time-delay transformer (STDT) connected between the ST and the NBS and with a snubber, which is in the arc and filament power leads, in connection with a spark gap. Alternatively, STDT's can be used for the HV and for the arc and filament power leads. This paper presents design details and experimental results of the overcurrent protection circuits

  3. Particle and energy transport studies on TFTR and implications for helium ash in future fusion devices

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Efthimion, P.C.; Rewoldt, G.; Stratton, B.C.; Tang, W.M.; Bell, R.E.; Grek, B.; Hulse, R.A.; Johnson, D.W.; Hill, K.W.; Mansfield, D.K.; McCune, D.; Mikkelsen, D.R.; Park, H.K.; Ramsey, A.T.; Scott, S.D.; Taylor, G.; Timberlake, J.; Zarnstorff, M.C.

    1992-01-01

    Particle and energy transport in tokamak plasmas have long been subjects of vigorous investigation. Present-day measurement techniques permit radially resolved studies of the transport of electron perturbations, low- and high-Z impurities, and energy. In addition, developments in transport theory provide tools that can be brought to bear on transport issues. Here, we examine local particle transport measurements of electrons, fully-stripped thermal helium, and helium-like iron in balanced-injection L-mode and enhanced confinement deuterium plasmas on TFTR of the same plasma current, toroidal field, and auxiliary heating power. He 2+ and Fe 24+ transport has been studied with charge exchange recombination spectroscopy, while electron transport has been studied by analyzing the perturbed electron flux following the same helium puff used for the He 2+ studies. By examining the electron and He 2+ responses following the same gas puff in the same plasmas, an unambiguous comparison of the transport of the two species has been made. The local energy transport has been examined with power balance analysis, allowing for comparisons to the local thermal fluxes. Some particle and energy transport results from the Supershot have been compared to a transport model based on a quasilinear picture of electrostatic toroidal drift-type microinstabilities. Finally, implications for future fusion reactors of the observed correlation between thermal transport and helium particle transport is discussed

  4. Periscope-camera system for visible and infrared imaging diagnostics on TFTR

    International Nuclear Information System (INIS)

    Medley, S.S.; Dimock, D.L.; Hayes, S.; Long, D.; Lowrence, J.L.; Mastrocola, V.; Renda, G.; Ulrickson, M.; Young, K.M.

    1985-05-01

    An optical diagnostic consisting of a periscope which relays images of the torus interior to an array of cameras is used on the Tokamak Fusion Test Reactor (TFTR) to view plasma discharge phenomena and inspect vacuum vessel internal structures in both visible and near-infrared wavelength regions. Three periscopes view through 20-cm-diameter fused-silica windows which are spaced around the torus midplane to provide a viewing coverage of approximately 75% of the vacuum vessel internal surface area. The periscopes have f/8 optics and motor-driven controls for focusing, magnification selection (5 0 , 20 0 , and 60 0 field of view), elevation and azimuth setting, mast rotation, filter selection, iris aperture, and viewing port selection. The four viewing ports on each periscope are equipped with multiple imaging devices which include: (1) an inspection eyepiece, (2) standard (RCA TC2900) and fast (RETICON) framing rate television cameras, (3) a PtSi CCD infrared imaging camera, (4) a 35 mm Nikon F3 still camera, or (5) a 16 mm Locam II movie camera with variable framing up to 500 fps. Operation of the periscope-camera system is controlled either locally or remotely through a computer-CAMAC interface. A description of the equipment and examples of its application are presented

  5. Periscope-camera system for visible and infrared imaging diagnostics on TFTR

    International Nuclear Information System (INIS)

    Medley, S.S.; Dimock, D.L.; Hayes, S.; Long, D.; Lowrance, J.L.; Mastrocola, V.; Renda, G.; Ulrickson, M.; Young, K.M.

    1985-01-01

    An optical diagnostic consisting of a periscope which relays images of the torus interior to an array of cameras is used on the Tokamak Fusion Test Reactor (TFTR) to view plasma discharge phenomena and inspect the vacuum vessel internal structures in both the visible and near-infrared wavelength regions. Three periscopes view through 20-cm-diam fused-silica windows which are spaced around the torus midplane to provide a viewing coverage of approximately 75% of the vacuum vessel internal surface area. The periscopes have f/8 optics and motor-driven controls for focusing, magnification selection (5 0 , 20 0 , and 60 0 field of view), elevation and azimuth setting, mast rotation, filter selection, iris aperture, and viewing port selection. The four viewing ports on each periscope are equipped with multiple imaging devices which include: (1) an inspection eyepiece, (2) standard (RCA TC2900) and fast (RETICON) framing rate television cameras, (3) a PtSi CCD infrared imaging camera, (4) a 35-mm Nikon F3 still camera, or (5) a 16-mm Locam II movie camera with variable framing rate up to 500 fps. Operation of the periscope-camera system is controlled either locally or remotely through a computer-CAMAC interface. A description of the equipment and examples of its application are presented

  6. Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code

    Directory of Open Access Journals (Sweden)

    Yutthapong Pinanroj

    2014-04-01

    Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.

  7. Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

    International Nuclear Information System (INIS)

    Anderson, J.L.; Gentile, C.; Kalish, M.; Kamperschroer, J.; Kozub, T.; LaMarche, P.; Murray, H.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Sissingh, R.; Swanson, J.; Tulipano, F.; Viola, M.; Voorhees, D.; Walters, R.T.

    1995-01-01

    The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.)

  8. Microinstability-based models for confinement properties and ignition criteria in tokamaks

    International Nuclear Information System (INIS)

    Tang, W.M.; Bishop, C.M.; Coppi, B.; Kaye, S.M.; Perkins, F.W.; Redi, M.H.; Rewoldt, G.

    1987-02-01

    This paper reports on results of theoretical studies dealing with: (1) the use of microinstability-based thermal transport models to interpret the anomalous confinement properties observed in key tokamak experiments such as TFTR and (2) the likely consequences of the presence of such instabilities for future ignition devices. Transport code simulations using profile-consistent forms of anomalous thermal diffusivities due to drift-type instabilities have yielded good agreement with the confinement times and temperatures observed in TFTR under a large variety of operating conditions including pellet-fuelling in both ohmic- and neutral-beam-heated discharges. With regard to achieving an optimal ignition margin, the adverse temperature scaling of anomalous losses caused by drift modes leads to the conclusion that it is best to operate at the maximum allowable density while holding the temperature close to the minimum value required for ignition

  9. Plans for the CIT [Compact Ignition Tokamak] instrumentation and control system

    International Nuclear Information System (INIS)

    Preckshot, G.G.

    1987-01-01

    Extensive experience with previous fusion experiments (TFTR, MFTF-B and others) is driving the design of the Instrumentation and Control System (I and C) for the Compact Ignition Tokamak (CIT) to be built at Princeton. The new design will reuse much equipment from TFTR and will be subdivided into six major parts: machine control, machine data acquisition, plasma diagnostic instrument control and instrument data acquisition, the database, shot sequencing and safety interlocks. In a major departure from previous fusion experiment control systems, the CIT machine control system will be a commercial process control system. Since the machine control system will be purchased as a completely functional product, we will be able to concentrate development manpower in plasma diagnostic instrument control, data acquisition, data processing and analysis, and database systems. We will discuss the issues driving the design, give a design overview and state the requirements upon any prospective commercial process control system

  10. Charged fusion product and fast ion loss in TFTR

    International Nuclear Information System (INIS)

    Zweben, S.J.; Darrow, D.S.; Fredrickson, E.D.; Mynick, H.E.; White, R.B.; Biglari, H.; Bretz, N.; Budny, R.; Bush, C.E.; Chang, C.S.; Chen, L.; Cheng, C.Z.; Fu, G.Y.; Hammett, G.W.; Hawryluk, R.J.; Hosea, J.; Johnson, L.; Mansfield, D.; McGuire, K.; Medley, S.S.; Nazikian, R.; Owens, D.K.; Park, H.; Park, J.; Phillips, C.K.; Schivell, J.; Stratton, B.C.; Ulrickson, M.; Wilson, R.; Young, K.M.; Fisher, R.; McChesney, J.; Fonck, R.; McKee, G.; Tuszewski, M.

    1993-03-01

    Several different fusion product and fast ion loss processes have been observed in TFTR using an array of pitch angle, energy and time resolved scintillator detectors located near the vessel wall. For D-D fusion products (3 MeV protons and 1 MeV tritons) the observed loss is generally consistent with expected first-orbit loss for Ip I MA. However, at higher currents, Ip = 1.4--2.5 MA, an NM induced D-D fusion product loss can be up to 3-4 times larger than the first-orbit loss, particularly at high beam powers, P ≥ 25 MW. The MHD induced loss of 100 KeV neutron beam ions and ∼0.5 MeV ICRF minority tail tons has also been measured ≤ 459 below the outer midplane. be potential implications of these results for D-T alpha particle experiments in TFTR and ITER are described

  11. Studies of tritiated co-deposited Layers in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Ascione, G.; Carpe, A.; Causey, R.A.; Hayashi, T.; Hogan, J.; Langish, S.W.; Nishi, M.F.; Shu, W.M.; Wampler, W.R.; Young, K.M.

    2000-01-01

    Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons, a stainless steel shutter and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.56 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling

  12. Studies of tritiated co-deposited layers in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Ascione, G.; Carpe, A.; Causey, R.A.; Hayashi, T.; Hogan, J.; Langish, S.; Nishi, M.; Shu, W.M.; Wampler, W.R.; Young, K.M.

    2000-01-01

    Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons, a stainless steel shutter and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.56 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling

  13. Studies of tritiated co-deposited layers in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Ascione, G.; Causey, R.A.; Hayaski, T.; Hogan, J.; Nishi, M.; Shu, W.M.; Wampler, William R.; Young, K.M.

    2000-01-01

    Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.5 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition

  14. Expansion of the TFTR neutral beam computer system

    International Nuclear Information System (INIS)

    McEnerney, J.; Chu, J.; Davis, S.; Fitzwater, J.; Fleming, G.; Funk, P.; Hirsch, J.; Lagin, L.; Locasak, V.; Randerson, L.; Schechtman, N.; Silber, K.; Skelly, G.; Stark, W.

    1992-01-01

    Previous TFTR Neutral Beam computing support was based primarily on an Encore Concept 32/8750 computer within the TFTR Central Instrumentation, Control and Data Acquisition System (CICADA). The resources of this machine were 90% utilized during a 2.5 minute duty cycle. Both interactive and automatic processes were supported, with interactive response suffering at lower priority. Further, there were additional computing requirements and no cost effective path for expansion within the Encore framework. Two elements provided a solution to these problems: improved price performance for computing and a high speed bus link to the SELBUS. The purchase of a Sun SPARCstation and a VME/SELBUS bus link, allowed offloading the automatic processing to the workstation. This paper describes the details of the system including the performance of the bus link and Sun SPARCstation, raw data acquisition and data server functions, application software conversion issues, and experiences with the UNIX operating system in the mixed platform environment

  15. ICRF sawtooth stabilization: Application on TFTR and CIT

    International Nuclear Information System (INIS)

    Hosea, J.C.; Phillips, C.K.; Stevens, J.E.; Wilson, J.R.; Bell, M.; Boivin, R.; Cavallo, A.; Colestock, P.; Fredrickson, E.; Hammett, G.; Hsuan, H.; Janos, A.; Jassby, D.; Jobes, F.; McGuire, K.; Mueller, D.; Nagayama, Y.; Owens, K.; Park, H.; Schmidt, G.; Stratton, B.; Taylor, G.; Wong, K.L.; Zweben, S.

    1991-03-01

    The use of ICRF heating to stabilize the core plasma sawtooth relaxations has been extended to TFTR where such stabilization has been produced at relatively low power in the L Mode regime at moderate density (P RF = 4 MW, 2.6 MW in helium and deuterium discharges, respectively, for the minority hydrogen ICRF heating regime with bar n e ∼2.5 x 10 13 cm -3 ). These results, as in the case of those obtained on JET, are qualitatively consistent with energetic ion stabilization of the m = 1, n = 1 ideal/resistive kink mode. The relevance of sawtooth stabilization to the primary regimes of interest on TFTR -- the high-Q supershot regime and the high density pellet injection regimes -- and on CIT -- the high density ICRF heated regime -- is considered in the context of the present theory and the projected ICRF power deposition characteristics. 35 refs., 11 figs

  16. A review of carbon blooms on JET and TFTR

    International Nuclear Information System (INIS)

    Ulrickson, M.

    1990-01-01

    Operation of JET and TFTR at high auxiliary heating power has resulted in the occurrence of phenomena called carbon blooms. The carbon bloom is characterized by a rapid increases in the emission of carbon spectral lines, the Z eff , the radiated power, and the plasma density. There is also a concurrent decrease in the neutron emission rate, stored energy, and plasma pressure. On both machines the source of the carbon is observed to be at localized (both toroidally and polidally) hot spots on either the divertor plates or limiters. The localized hot spots are due to one or more of the following: disruption damage spots, misalignment of tiles, and/or exposed edges of tiles. The occurrence of carbon blooms limits the performance of the highest input power plasmas on both machines. This paper reviews the carbon bloom phenomenon as it occurs on both JET and TFTR. (orig.)

  17. Power and particle balance during neutral beam injection in TFTR

    International Nuclear Information System (INIS)

    Pitcher, C.S.; Budny, R.V.; Hill, K.W.; Kilpatrick, S.J.; Manos, D.M.; Medley, S.S.; Ramsey, A.T.

    1991-05-01

    Detailed boundary plasma measurements on TFTR have been made during a NBI power scan in the range P tot = 1MW--20MW in the L-mode regime. The behavior of the plasma density left-angle n e right-angle, radiated power P rad , carbon and deuterium fluxes Γ C , Γ D , and Ζ eff can be summarized as, left-angle n e right-angle ∝ P tot 1/2 , P rad , Γ C , Γ D ∝ P tot , and Ζ eff ∼ constant. It is shown that central fuelling by the neutral beams plays a minor role in the particle balance of the discharge. More important is the NBI role in the power balance. The TFTR data during NBI originate primarily at the graphite limiter

  18. Heat pulse propagation studies on DIII-D and the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Fredrickson, E. D.; Austin, M. E.; Groebner, R.; Manickam, J.; Rice, B.; Schmidt, G.; Snider, R.

    2000-12-01

    Sawtooth phenomena have been studied on DIII-D and the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24]. In the experiments the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. For the first time, measurements of a strong ballistic electron heat pulse were made in a shaped tokamak (DIII-D) [J. Luxon and DIII-D Group, in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] and the "ballistic effect" was stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the fast growth phase of the sawtooth precursor. Fast, 2 ms interval, measurements on DIII-D were made of the ion temperature evolution following sawteeth and partial sawteeth to document the ion heat pulse characteristics. It is found that the ion heat pulse does not exhibit the very fast, "ballistic" behavior seen for the electrons. Further, for the first time it is shown that the electron heat pulses from partial sawtooth crashes (on DIII-D and TFTR) are seen to propagate at speeds close to those expected from the power balance calculations of the thermal diffusivities whereas heat pulses from fishbones propagate at rates more consistent with sawtooth induced heat pulses. These results suggest that the fast propagation of sawtooth-induced heat pulses is not a feature of nonlinear transport models, but that magnetohydrodynamic events can have a strong effect on electron thermal transport.

  19. Heating and transport in TFTR D-T plasmas

    International Nuclear Information System (INIS)

    Zarnstorff, M.C.; Scott, S.D.

    1994-01-01

    The confinement and heating of supershot plasmas are significantly enhanced with tritium beam injection relative to deuterium injection in TFTR. The global energy confinement and local thermal transport are analyzed for deuterium and tritium fueled plasmas to quantify their dependence on the average mass of the hydrogenic ions. The radial profiles of the deuterium and tritium densities are determined from the DT fusion neutron emission profile

  20. Automatic and manual operation modes of the TFTR maintenance manipulator

    International Nuclear Information System (INIS)

    Boehme, G.; Gumb, L.; Lotz, E.; Mueller, G.; Selig, M.

    1987-01-01

    The remote in-vessel operations scheduled to maintain the TFTR at Princeton, NJ, USA, comprise inspection, calibration, cleaning and protective tile replacement. The environmental conditions inside the torus vessel are ultra high vacuum, moderate γ-radiation and 150 0 C temperature of the vessel structure. The Princeton Plasma Physics Laboratory (PPPL) and KfK are jointly developing a maintenance manipulator (MM) which can perform these tasks. (orig./HP)

  1. Observation of neoclassical transport in reverse shear plasmas on TFTR

    International Nuclear Information System (INIS)

    Efthimion, P.C.; Goeler, S. von; Houlberg, W.A.

    1999-01-01

    Perturbative experiments on TFTR have investigated the transport of multiple ion species in reverse shear (RS) plasmas. The profile evolutions of trace tritium and helium and intrinsic carbon indicate the formation of core particle transport barriers in enhanced reverse shear (ERS) plasmas. There is an order of magnitude reduction in the particle diffusivity inside the RS region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)

  2. Compound sawtooth study in ohmically heated TFTR plasmas

    International Nuclear Information System (INIS)

    Yamada, H.; McGuire, K.; Colchin, D.

    1985-09-01

    Compound sawtooth activity has been observed in ohmically heated, high current, high density TFTR plasmas. Commonly called ''double sawteeth,'' such sequences consist of a repetitive series of subordinate relaxations followed by a main relaxation with a different inversion radius. The period of such compound sawteeth can be as long as 100 msec. In other cases, however, no compound sawteeth or bursts of them can be observed in discharges with essentially the same parameters

  3. Structural analysis and optimization procedure of the TFTR device substructure

    International Nuclear Information System (INIS)

    Driesen, G.

    1975-10-01

    A structural evaluation of the TFTR device substructure is performed in order to verify the feasibility of the proposed design concept as well as to establish a design optimization procedure for minimizing the material and fabrication cost of the substructure members. A preliminary evaluation of the seismic capability is also presented. The design concept on which the analysis is based is consistent with that described in the Conceptual Design Status Briefing report dated June 18, 1975

  4. Extension of TFTR operations to higher toroidal field levels

    International Nuclear Information System (INIS)

    Woolley, R.D.

    1995-01-01

    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 1994 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

  5. Design of a tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

    Milora, S.L.; Gouge, M.J.; Fisher, P.W.; Combs, S.K.; Cole, M.J.; Wysor, R.B.; Fehling, D.T.; Foust, C.R.; Baylor, L.R.; Schmidt, G.L.; Barnes, G.W.; Persing, R.G.

    1991-01-01

    The TFTR tritium pellet injector (TPI) is designed to provide a tritium pellet fueling capability with pellet speeds in the 1- to 3 km/s-range for the TFTR D-T phase. The existing TFTR deuterium pellet injector is being modified at Oak Ridge National Laboratory to provide a fourshot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns a two -stage light gas gun driver. The pipe gun concept has been qualified for tritium operation by the tritium proof-of-principle injector experiments conducted on the Tritium Systems Test Assembly at Los Alamos National Laboratory. In these experiments, tritium and D-T pellets were accelerated to speeds near 1.5 km/s. The TPI is being designed for pellet sizes in the range from 3.43 to 4.0 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation will be controlled by a programmable logic controller. 7 refs., 4 figs

  6. Development of the TFTR neutral beam injection system

    International Nuclear Information System (INIS)

    Prichard, B.A. Jr.

    1977-01-01

    The TFTR Neutral Beam Lines are designed to inject 20 MW of 120 keV neutral deuterium atoms into the plasma. This is accomplished using 12 sources, 65 amperes each, mounted in 4 beam lines. The 120 kV sources and a prototype beam line are being developed. The implementation of these beam lines has required the development of several associated pieces of hardware. 200 kV switch tubes for the power supplies are being developed for modulation and regulation of the accelerating supplies. A 90 cm metallic seal gate valve capable of sealing against atmosphere in either direction is being developed for separating the torus and beam line vacuum systems. A 70 x 80 cm fast shutter valve is also being developed to limit tritium migration from the torus into the beam line. Internal to the beam line a calorimeter, ion dump and deflection magnet have been designed to handle three beams, and optical diagnostics utilizing the doppler broadening and doppler shift of light emitted from the accelerated beam are being developed. The control and monitoring of the 12 sources will be done via the TFTR computer control system (CICADA) as will other parts of the machine, and software is being developed to condition and operate the sources automatically. The prototype beam line is scheduled to begin operation in the fall of 1978 and all four production beam lines on TFTR in 1982

  7. Automatic generation of computer programs servicing TFTR console displays

    International Nuclear Information System (INIS)

    Eisenberg, H.

    1983-01-01

    A number of alternatives were considered in providing programs to support the several hundred displays required for control and monitoring of TFTR equipment. Since similar functions were performed, an automated method of creating programs was suggested. The complexity of a single program servicing as many as thirty consoles mitigated against that approach. Similarly, creation of a syntactic language while elegant, was deemed to be too time consuming, and had the disadvantage of requiring a working knowledge of the language on a programming level. It was elected to pursue a method of generating an individual program to service a particular display. A feasibility study was conducted and the Control and Monitor Display Generator system (CMDG) was developed. A Control and Monitor Display Service Program (CMDS) provides a means of performing monitor and control functions for devices associated with TFTR subsystems, as well as other user functions, via TFTR Control Consoles. This paper discusses the specific capabilities provided by CMDS in a usage context, as well as the mechanics of implementation

  8. Varennes Tokamak

    International Nuclear Information System (INIS)

    Cumyn, P.B.

    A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

  9. Overview of the first workshop on alpha particle physics in TFTR

    International Nuclear Information System (INIS)

    Zweben, S.J.; Biglari, H.

    1991-07-01

    The ''First Workshop on Alpha Physics in TFTR'' was held at the Princeton Plasma Physics Lab March 28--29, 1991. The motivation for this meeting was to clarify and strengthen the TFTR alpha physics program, and to increase the involvement of the fusion community outside PPPL in the TFTR D-T experiments. Therefore the meeting was sharply focused on alpha physics relevant to the upcoming TFTR D-T simulation, and was asked to devote half of his talk to specific TFTR issues. The Workshop consisted of 27 talks on: (1) experimental possibilities; (2) theoretical possibilities; (3) diagnostic possibilities; (4) relevance for future machines; and (5) discussion/summary session. This summary contains a brief sampling of the new results and ideas brought out by these talks, followed by two more general overviews of the status of experiment and theory

  10. Tokamak physics

    International Nuclear Information System (INIS)

    Haines, M.G.

    1984-01-01

    The physical conditions required for breakeven in thermonuclear fusion are derived, and the early conceptual ideas of magnetic confinement and subsequent development are followed, leading to present-day large scale tokamak experiments. Confinement and diffusion are developed in terms of particle orbits, whilst magnetohydrodynamic stability is discussed from energy considerations. From these ideas are derived the scaling laws that determine the physical size and parameters of this fusion configuration. It becomes clear that additional heating is required. However there are currently several major gaps in our understanding of experiments; the causes of anomalous electron energy loss and the major current disruption, the absence of the 'bootstrap' current and what physics determines the maximum plasma pressure consistent with stability. The understanding of these phenomena is a major challenge to plasma physicists. (author)

  11. In-situ Tritium Measurements of the Tokamak Fusion Test Reactor Bumper Limiter Tiles Post D-T Operations

    International Nuclear Information System (INIS)

    C.A. Gentile; C.H. Skinner; K.M. Young; M. Nishi; S. Langish; et al

    1999-01-01

    The Princeton Plasma Physics Laboratory (PPPL) Engineering and Research Staff in collaboration with members of the Japan Atomic Energy Research Institute (JAERI), Tritium Engineering Laboratory have commenced in-situ tritium measurements of the TFTR bumper limiter. The Tokamak Fusion Test Reactor (TFTR) operated with tritium from 1993 to 1997. During this time ∼ 53,000 Ci of tritium was injected into the TFTR vacuum vessel. After the cessation of TFTR plasma operations in April 1997 an aggressive tritium cleanup campaign lasting ∼ 3 months was initiated. The TFTR vacuum vessel was subjected to a regimen of glow discharge cleaning (GDC) and dry nitrogen and ''moist air'' purges. Currently ∼ 7,500 Ci of tritium remains in the vacuum vessel largely contained in the limiter tiles. The TFTR limiter is composed of 1,920 carbon tiles with an average weight of ∼ 600 grams each. The location and distribution of tritium on the TFTR carbon tiles are of considerable interest. Future magnetically confined fusion devices employing carbon as a limiter material may be considerably constrained due to potentially large tritium inventories being tenaciously held on the surface of the tiles. In-situ tritium measurements were conducted in TFTR bay L during August and November 1998. During the bay L measurement campaign open wall ion chambers and ultra thin thermoluminscent dosimeters (TLD) affixed to a boom and end effector were deployed into the vacuum vessel. The detectors were designed to make contact with the surface of the bumper limiter tile and to provide either real time (ion chamber) or passive (TLD) indication of the surface tritium concentration. The open wall ion chambers were positioned onto the surface of the tile in a manner which employed the surface of the tile as one of the walls of the chamber. The ion chambers, which are (electrically) gamma insensitive, were landed at four positions per tile. The geometry for landing the TLD's provided measurement at 24

  12. High poloidal beta equilibria in TFTR limited by a natural inboard poloidal field null

    International Nuclear Information System (INIS)

    Sabbagh, S.A.; Gross, R.A.; Mauel, M.E.; Navratil, G.A.; Bell, M.G.; Bell, R.; Bitter, M.; Bretz, N.L.; Budny, R.V.; Bush, C.E.; Chance, M.S.; Efthimion, P.C.; Fredrickson, E.D.; Hatcher, R.; Hawryluk, R.J.; Hirshman, S.P.; Janos, A.C.; Jardin, S.C.; Jassby, D.L.; Manickam, J.; McCune, D.C.; McGuire, K.M.; Medley, S.S.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Okabayashi, M.; Park, H.K.; Ramsey, A.T.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Wieland, R.M.; Zarnstorff, M.C.; Kesner, J.; Marmar, E.S.; Terry, J.L.

    1991-07-01

    Recent operation of the Tokamak Fusion Test Reactor TFTR, has produced plasma equilibria with values of Λ triple-bond β p eq + l i /2 as large as 7, εβ p dia triple-bond 2μ 0 ε /much-lt B p much-gt 2 as large as 1.6, and Troyon normalized diamagnetic beta, β N dia triple-bond 10 8 t perpendicular>aB 0 /I p as large as 4.7. When εβ p dia approx-gt 1.25, a separatrix entered the vacuum chamber, producing a naturally diverted discharge which was sustained for many energy confinement times, τ E . The largest values of εβ p and plasma stored energy were obtained when the plasma current was ramped down prior to neutral beam injection. The measured peak ion and electron temperatures were as large as 24 keV and 8.5 keV, respectively. Plasma stored energy in excess of 2.5 MJ and τ E greater than 130 msec were obtained. Confinement times of greater than 3 times that expected from L-mode predictions have been achieved. The fusion power gain. Q DD , reached a values of 1.3 x 10 -3 in a discharge with I p = 1 MA and εβ p dia = 0.85. A large, sustained negative loop voltage during the steady state portion of the discharge indicates that a substantial non-inductive component of I p exists in these plasmas. Transport code analysis indicates that the bootstrap current constitutes up to 65% of I p . Magnetohydrodynamic (MHD) ballooning stability analysis shows that while these plasmas are near, or at the β p limit, the pressure gradient in the plasma core is in the first region of stability to high-n modes. 24 refs., 10 figs

  13. ICRF heating and transport of deuterium-tritium plasmas in TFTR

    International Nuclear Information System (INIS)

    Murakami, M.; Batchelor, D.B.; Bush, C.E.

    1994-01-01

    This paper describes results of the first experiments utilizing high-power ion cyclotron range of frequency (ICRF) to heat deuterium-tritium (D-T) plasmas in reactor-relevant regimes on the Tokamak Fusion Test Reactor (TFTR). Results from these experiments have demonstrated efficient core, second harmonic, tritium heating of D-T supershot plasmas with tritium concentrations ranging from 6%--40%. Significant direct ion heating on the order of 60% of the input radio frequency (rf) power has been observed. The measured deposition profiles are in good agreement with two-dimensional modeling code predictions. Confinement in an rf-heated supershot is at least similar to that without rf, and possibly better in the electron channel. Efficient electron heating via mode conversion of fast waves to ion Bernstein waves (IBW) has been demonstrated in ohmic, deuterium-deuterium and DT-neutral beam injection plasmas with high concentrations of minority 3 He (n 3 He /n e > 10%). By changing the 3 He concentration or the toroidal field strength, the location of the mode-conversion radius was varied. The power deposition profile measured with rf power modulation showed that up to 70% of the power can be deposited on electrons at an off-axis position. Preliminary results with up to 4 MW coupled into the plasma by 90-degree phased antennas showed directional propagation of the mode-converted IBW. Heat wave propagation showed no strong inward thermal pinch in off-axis heating of an ohmically-heated (OH) target plasma in TFIR

  14. Radio frequency plasma heating in large tokamak systems near the lower hybrid resonance

    International Nuclear Information System (INIS)

    Deitz, A.; Hooke, W.M.

    1975-01-01

    The frequency range, power, efficiency, and pulse length of a high power rf system are discussed as they might be applied to the TFTR Tokamak facility as well as on a full scale reactor. Comparisons are made of the size, power output, and costs to obtain microwave power sufficient to satisfy the physics requirements. A new microwave feed concept is discussed which will improve the coupling of the microwave energy into the plasma. The unique advantages of waveguide feed systems is apparent when one considers the practical problems associated with coupling supplementary heating energy into a reactor

  15. Effect of alpha drift and instabilities on tokamak plasma edge conditions

    International Nuclear Information System (INIS)

    Miley, G.H.; Choi, C.K.

    1983-01-01

    As suprathermal fusion products slow down in a Tokamak, their average drift is inward. The effect of this drift on the alpha heating and thermalization profiles is examined. In smaller TFTR-type devices, heating in the outer region can be cut in half. Also, the fusion-product energy-distribution near the plasma edge has a positive slope with increasing energy, representing a possible driving mechanism for micro-instabilities. Another instability that can seriously affect outer plasma conditions and shear Alfven transport of alphas is also considered

  16. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    M.E. Lumia; C.A. Gentile

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed

  17. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    CERN Document Server

    Lumia, M E

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  18. Escaping 1 MeV tritons in TFTR

    International Nuclear Information System (INIS)

    Zweben, S.J.; Strachan, J.D.; Boivin, R.; Cavallo, A.; Fredrickson, E.D.; McGuire, K.; Mynick, H.E.; White, R.B.

    1989-01-01

    1 MeV tritons created by D-D reactions can simulate the 'single-particle' behavior expected with 3.5 MeV D-T alphas, since the gyroradii and slowing-down of these two particles are similar. This paper describes measurements of the flux of escaping 1 MeV tritons from the TFTR plasma during high power D 0 →D neutral beam injection, and shows that in most cases the observed triton loss is consistent with the classical (single-particle) first-orbit loss model. In this model tritons are lost if their first orbit intersects the wall due to their large banana width, while almost all tritons confined on their first orbit should stay confined until thermalized. The triton detectors are ZnS(Ag) scintillator screens housed in light-tight boxes located just outside the plasma boundary at the bottom of the TFTR vessel. They are particle 'pinhole' cameras which can resolve the triton flux vs. pitch angle (to ±5 o ), energy (to ±50 %), and time (to <20 μsec). The 2-D images of triton flux onto these scintillators are optically coupled to either an intensified TV camera or to photomultiplyer tubes for fast time resolution. The soft x-ray background in an earlier prototype has been eliminated. Although there are presently 8 such detectors in TFTR, this paper discusses results from only the detector located just below the vessel center (R=259 cm, r=102 cm). Note that the '1 MeV triton' signal discussed below also has about a 30 % contribution from 3 MeV protons; however, since these two particles have identical gyroradii they should behave alike. 5 refs., 5 figs

  19. Status of tokamak research

    International Nuclear Information System (INIS)

    Rawls, J.M.

    1979-10-01

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

  20. Wall conditioning experiments on TFTR using impurity pellet injection

    International Nuclear Information System (INIS)

    Strachan, J.D.; Mansfield, D.K.; Bell, M.G.; Collins, J.; Ernst, D.; Hill, K.; Hosea, J.; Timberlake, J.; Ulrickson, M.; Terry, J.; Marmar, E.; Snipes, J.

    1994-01-01

    This work describes experiments intended to optimize the limiter conditioning for TFTR supershots. It is shown that deposition of thin layers of lithium on the limiters by impurity pellet injection changes the plasma-wall interaction and improves supershot performance. Series of up to ten Ohmic plasmas each with two lithium pellets were useful in pre-conditioning the limiter. Generally, plasma performance increased with the amount of lithium deposited up to the maximal amount which could be deposited. Experiments were performed with different materials being deposited (carbon, boron and lithium) and with different methods of deposition. ((orig.))

  1. Overview of the TFTR Lithium Blanket Module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

    The LBM (Lithium Blanket Module) is an approximately cubic module, about 80 cm on each side, with construction representative of a helium-cooled lithium oxide fusion reactor blanket module. Measurements of neutron transport and tritium breeding in the LBM will be made in irradiation programs first with a point-neutron source, and subsequently with the D-D and D-T fusion-neutron sources of the TFTR. This paper summarizes the objectives of the LBM program, the design, development and construction of the LBM, and progress in the experimental tests

  2. Alpha particle loss in the TFTR DT experiments

    International Nuclear Information System (INIS)

    Zweben, S.J.; Darrow, D.S.; Herrmann, H.W.

    1995-01-01

    Alpha particle loss was measured during the TFTR DT experiments using a scintillator detector located at the vessel bottom in the ion grad-B drift direction. The DT alpha particle loss to this detector was consistent with the calculated first-orbit loss over the whole range of plasma current I=0.6-2.7 MA. In particular, the alpha particle loss rate per DT neutron did not increase significantly with fusion power up to 10.7 MW, indicating the absence of any new ''collective'' alpha particle loss processes in these experiments

  3. ICRF-induced DD fusion product losses in TFTR

    International Nuclear Information System (INIS)

    Darrow, D.S.; Zweben, S.J.; Budny, R.V.

    1994-10-01

    When ICRF power is applied to TFTR plasmas in which there is no externally-supplied minority species, an enhanced loss of DD fusion products results. The characteristics of the loss are consistent with particles at or near the birth energy having their perpendicular velocity increased by the ICRF such that those near the passing/trapped boundary are carried into the first orbit loss cone. A rudimentary model of this process predicts losses of a magnitude similar to those seen. Extrapolations based upon this data for hypothetical ICRF ash removal from reactor plasmas suggest that the technique will not be energy efficient

  4. TAE Saturation of Alpha Particle Driven Instability in TFTR

    International Nuclear Information System (INIS)

    Berk, H.L.; Chen, Y.; Gorelenkov, N.N.; White, R.B.

    1998-01-01

    A nonlinear theory of kinetic instabilities near threshold [H.L. Berk, et al., Plasma Phys. Rep. 23, (1997) 842] is applied to calculate the saturation level of Toroidicity-induced Alfvn Eigenmodes (TAE) and be compared with the predictions of (delta)f method calculations [Y. Chen, Ph.D. Thesis, Princeton University, 1998]. Good agreement is observed between the predictions of both methods and the predicted saturation levels are comparable with experimentally measured amplitudes of the TAE oscillations in TFTR [D.J. Grove and D.M. Meade, Nucl. Fusion 25, (1985) 1167

  5. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    International Nuclear Information System (INIS)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to ∼9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS ∼6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored

  6. New tritium monitor for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Jalbert, R.A.

    1985-01-01

    At DT-fueled fusion reactors, there will be a need for tritium monitors that can simultaneously measure in real time the concentrations of HTO, HT and the activated air produced by fusion neutrons. Such a monitor has been developed, tested and delivered to the Princeton Plasma Physics Laboratory for use at the Tokamak Fusion Test Reactor (TFTR). It uses semipermeable membranes to achieve the removal of HTO from the sampled air for monitoring and a catalyst to convert the HT to HTO, also for removal and monitoring. The remaining air, devoid of tritium, is routed to a third detector for monitoring the activated air. The sensitivities are those that would be expected from tritium instruments employing conventional flow-through ionization chambers: 1 to 3 μCi/m 3 . Its discriminating ability is approximately 10 -3 for any of the three components (HTO, HT and activated air) in any of the other two channels. For instance, the concentration of HT in the HTO channel is 10 -3 times its original concentration in the sampled air. This will meet the needs of TFTR

  7. Decontamination and decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Walton, G.R.; Perry, E.D.; Commander, J.C.; Spampinato, P.T.

    1994-01-01

    The Tokamak Fusion Test Reactor (TFTR) is scheduled to complete its end-of-life deuterium-tritium (D-T) experiments in September 1994. The D-T operation will result in the TFTR machine structure becoming activated, and plasma facing and vacuum components will be contaminated with tritium. The resulting machine activation levels after a two year cooldown period will allow hands on dismantling for external structures, but require remote dismantling for the vacuum vessel. The primary objective of the Decontamination and Decommissioning (D ampersand D) Project is to provide a facility for construction of a new Department of Energy (DOE) experimental fusion reactor by March 1998. The project schedule calls for a two year shutdown period when tritium decontamination of the vacuum vessel, neutral beam injectors and other components will occur. Shutdown will be followed by an 18 month period of D ampersand D operations. The technical objectives of the project are to: safely dismantle and remove components from the test cell complex; package disassembled components in accordance with applicable regulations; ship packages to a DOE approved disposal or material recycling site; and develop expertise using remote disassembly techniques on a large scale fusion facility. This paper discusses the D ampersand D objectives, the facility to be decommissioned, and the technical plan that will be implemented

  8. Stationary Flowing Liquid Lithium (SFLiLi) systems for tokamaks

    Science.gov (United States)

    Zakharov, Leonid; Gentile, Charles; Roquemore, Lane

    2013-10-01

    The present approach to magnetic fusion which relies on high recycling plasma-wall interaction has exhausted itself at the level of TFTR, JET, JT-60 devices with no realistic path to the burning plasma. Instead, magnetic fusion needs a return to its original idea of insulation of the plasma from the wall, which was the dominant approach in the 1970s and upon implementations has a clear path to the DEMO device with PDT ~= 100 MW and Qelectric > 1 . The SFLiLi systems of this talk is the technology tool for implementation of the guiding idea of magnetic fusion. It utilizes the unique properties of flowing LiLi to pump plasma particles and, thus, insulate plasma from the walls. The necessary flow rate, ~= 1 g3/s, is very small, thus, making the use of lithium practical and consistent with safety requirements. The talk describes how chemical activity of LiLi, which is the major technology challenge of using LiLi in tokamaks, is addressed by SFLiLi systems at the level of already performed (HT-7) experiment, and in ongoing implementations for a prototype of SFLiLi for tokamak divertors and the mid-plane limiter for EAST tokamak (to be tested in the next experimental campaign). This work is supported by US DoE contract No. DE-AC02-09-CH11466.

  9. Engineering parameters for four ignition TNS tokamak reactor systems

    International Nuclear Information System (INIS)

    Varljen, T.C.; Gibson, G.; French, J.W.; Heck, F.M.

    1977-01-01

    The ORNL/Westinghouse program for The Next Step (TNS) tokamak beyond TFTR has examined a large number of potential configurations for D-T burning ignition tokamak systems. An objective of this work has been to quantify the trade-offs associated with the assumption of certain plasma physics criteria and toroidal field coil technologies. Four tokamak system point designs are described, each representative of the TF coil technologies considered, to illustrate the engineering features associated with each concept. Point designs, such as the ones discussed herein, have been used to develop component size, performance and cost scaling relationships which have been incorporated in a digital computer code to facilitate an examination of the total design and cost impact of candidate design approaches. The point designs which are described are typical, however, they have not been individually optimized. The options are distinguished by the TF coil technology chosen and include: (1) a high field water-cooled copper TF system, (2) a moderate field NbTi superconducting TF system, (3) a high field Nb 3 Sn superconducting TF system, and (4) a high field hybrid TF system with outer NbTi superconducting windings and inner water-cooled copper windings. Descriptions are provided for the major device components and all major support systems including power supplies, vacuum systems, fuel systems, heat transport and facility systems

  10. Remote maintenance of in-vessel components in Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Loesser, G.D.; Heitzenroeder, P.; Kungl, D.; Dylla, H.F.; Cerdan, G.

    1990-01-01

    The Tokamak Fusion Test Reactor (TFTR) will generate a total of 3 x 10 21 neutrons during its planned D-T operational period. A maintenance manipulator has been designed and tested to minimize personnel radiation during in-vessel maintenance activities. Its functions include visual inspection, first-wall tile replacement, cleaning, diagnostics calibrations and leak detection. To meet these objectives, the TFTR maintenance manipulator is required to be operable in the TFTR high vacuum environment, typically -8 torr, ( -6 Pa). Geometrically, the manipulator must extend 180 0 in either direction around the torus to assure complete coverage of the vessel first-wall. The manipulator consists of a movable carriage, and movable articulated link sections which are driven by electrical actuators. The boom has vertical load capacity of 455 kg and lateral load capacity of 46 kg. The boom can either be fitted with a general inspection arm or dextrous slave arms. The general inspection arm is designed to hold the leak detector and an inspection camera; it is capable of rotation along two axes and has a linkage system which permits motion normal to the vacuum vessel wall. All systems except the dextrous slave arms are operable in a vacuum. (author)

  11. Design of deuterium and tritium pellet injector systems for Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Wysor, R.B.; Baylor, L.R.; Bryan, W.E.

    1985-01-01

    Three pellet injector designs developed by the Oak Ridge National Laboratory (ORNL) are planned for the Tokamak Fusion Test Reactor (TFTR) to reach the goal of a tritium pellet injector by 1988. These are the Repeating Pneumatic Injector (RPI), the Deuterium Pellet Injector (DPI) and the Tritium Pellet Injector (TPI). Each of the pellet injector designs have similar performance characteristics in that they deliver up to 4-mm-dia pellets at velocities up to 1500 m/s with a dsign goal to 2000 m/s. Similar techniques are utilized to freeze and extrude the pellet material. The injector systems incorporate three gun concepts which differ in the number of gun barrels and the method of forming and chambering the pellets. The RPI, a single barrel repeating design, has been operational on TFTR since April 1985. Fabrication and assembly are essentially complete for DPI, and TPI is presently on hold after completing about 80% of the design. The TFTR pellet injector program is described, and each of the injector systems is described briefly. Design details are discussed in other papers at this symposium

  12. Results from deuterium-tritium tokamak confinement experiments

    International Nuclear Information System (INIS)

    Hawryluk, R.J.

    1997-02-01

    Recent scientific and technical progress in magnetic fusion experiments has resulted in the achievement of plasma parameters (density and temperature) which enabled the production of significant bursts of fusion power from deuterium-tritium fuels and the first studies of the physics of burning plasmas. The key scientific issues in the reacting plasma core are plasma confinement, magnetohydrodynamic (MHD) stability, and the confinement and loss of energetic fusion products from the reacting fuel ions. Progress in the development of regimes of operation which have both good confinement and are MHD stable have enabled a broad study of burning plasma physics issues. A review of the technical and scientific results from the deuterium-tritium experiments on the Joint European Torus (JET) and the Tokamak Fusion Test Reactor (TFTR) is given with particular emphasis on alpha-particle physics issues

  13. Tritium pellet injector design for tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Fisher, P.W.; Baylor, L.R.; Bryan, W.E.

    1985-01-01

    A tritium pellet injector (TPI) system has been designed for the Tokamak Fusion Test Reactor (TFTR) Q approx. 1 phase of operation. The injector gun utilizes a radial design with eight independent barrels and a common extruder to minimize tritium inventory. The injection line contains guide tubes with intermediate vacuum pumping stations and fast valves to minimize propellant leakage to the torus. The vacuum system is designed for tritium compatibility. The entire injector system is contained in a glove box for secondary containment protection against tritium release. Failure modes and effects have been analyzed, and structural analysis has been performed for most intense predicted earthquake conditions. Details of the design and operation of this system are presented in this paper

  14. Experience with high heat flux components in large tokamaks

    International Nuclear Information System (INIS)

    Chappuis, P.; Dietz, K.J.; Ulrickson, M.

    1991-01-01

    The large present day tokamaks. i.e.JET, TFTR, JT-60, DIII-D and Tore Supra are machines capable of sustaining plasma currents of several million amperes. Pulse durations range from a few seconds up to a minute. These large machines have been in operation for several years and there exists wide experience with materials for plasma facing components. Bare and coated metals, bare and coated graphites and beryllium were used for walls, limiters and divertors. High heat flux components are mainly radiation cooled, but stationary cooling for long pulse duration is also employed. This paper summarizes the experience gained in the large machines with respect to material selection, component design, problem areas, and plasma performance. 2 tabs., 26 figs., 50 refs

  15. Conceptual designs of power tokamak-type thermonuclear reactors

    International Nuclear Information System (INIS)

    Shejndlin, A.E.; Nedospasov, A.V.

    1978-01-01

    Physico-technical and ecological aspects of conceptual designing power tokamak-type reactors have been briefly considered. Only ''pure'' (''non-hybride'') reactors are discussed. Presented are main plasma-physical parameters, characteristics of blankets and magnetic systems of the following projects: PPPL; V-2; V-3; Culham-2, JAERI; TBEh-2500; TFTR. Two systems of the first wall protection have been considered: divertor one and by means of a layer of a cool turbulent plasma. Examined are the following problems: fuel loading, choice of the first wall material, blanket structure, magnetic system, environmental contamination. The comparison of relative hazards of fast neutron reactors and fusion reactors has shown that in respect of fusion reactors the biological hazard potential value is less by one-two orders

  16. Hot spots on the neutralizer plates of a tokamak

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.

    1991-01-01

    The formation of hot spots on the neutralizing surfaces of tokamaks may be one of the reasons for the entry of large impurity fluxes into the plasmas of TFTR and JET (the so-called carbon catastrophe or carbon bloom) with high auxiliary heating powers. At this time it is unclear whether these hot spots are caused just by nonuniformities on the neutralizer surface or whether their appearance is the result of some more general behavior, with the surface nonuniformities only showing up as seed perturbations. In this paper it is shown that hot spots can also develop on smooth surfaces of carbon neutralizer plates as a result of the contraction of a heat flux incident on the plates

  17. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  18. Alpha diagnostics using pellet charge exchange: Results on TFTR and prospects for ITER

    International Nuclear Information System (INIS)

    Fisher, R.K.; Duong, H.H.; McChesney, J.M.

    1996-05-01

    Confinement of alpha particles is essential for fusion ignition and alpha physics studies are a major goal of the TFTR, JET, and ITER DT experiments, but alpha measurements remain one of the most challenging plasma diagnostic tasks. The Pellet Charge Exchange (PCX) diagnostic has successfully measured the radial density profile and energy distribution of fast (0.5 to 3.5 MeV) confined alpha particles in TFTR. This paper describes the diagnostic capabilities of PCX demonstrated on TFTR and discusses the prospects for applying this technique to ITER. Major issues on ITER include the pellet's perturbation to the plasma and obtaining satisfactory pellet penetration into the plasma

  19. Perspectives gained from ICRF physics studies on TFTR

    International Nuclear Information System (INIS)

    Phillips, C.K.; Bell, M.; Batha, S.

    1998-01-01

    The physics of ICRF heating and current drive has been studied on TFTR for over a decade. Following the early low power coupling studies, high power experiments resulted in sawtooth stabilization, the first observation of RF-driven excitation of toroidal Alfven eigenmodes, and the discovery of a mode conversion scenario for localized off-axis electron heating. The program culminated with the first studies of high power ICRF heating and profile control in tritium-rich high performance plasmas. A significant part of the concluding experiments centered on the potential of ICRF to drive sheared flows in order to suppress turbulence in the plasma core. Initial measurements taken with a novel poloidal velocity diagnostic suggest that localized sheared poloidal flows can be driven with ion Bernstein waves excited directly or else via mode conversion from a propagating fast magnetosonic wave. In this paper, recent results from TFTR on wave-based profile control techniques will be summarized along with suggestions for future studies elsewhere

  20. Modeling of high power ICRF heating experiments on TFTR

    International Nuclear Information System (INIS)

    Phillips, C.K.; Wilson, J.R.; Bell, M.; Fredrickson, E.; Hosea, J.C.; Majeski, R.; Ramsey, A.; Rogers, J.H.; Schilling, G.; Skinner, C.; Stevens, J.E.; Taylor, G.; Wong, K.L.; Murakami, M.

    1993-01-01

    Over the past two years, ICRF heating experiments have been performed on TFTR in the hydrogen minority heating regime with power levels reaching 11.2 MW in helium-4 majority plasmas and 8.4 MW in deuterium majority plasmas. For these power levels, the minority hydrogen ions, which comprise typically less than 10% of the total electron density, evolve into la very energetic, anisotropic non-Maxwellian distribution. Indeed, the excess perpendicular stored energy in these plasmas associated with the energetic minority tail ions is often as high as 25% of the total stored energy, as inferred from magnetic measurements. Enhanced losses of 0.5 MeV protons consistent with the presence of an energetic hydrogen component have also been observed. In ICRF heating experiments on JET at comparable and higher power levels and with similar parameters, it has been suggested that finite banana width effects have a noticeable effect on the ICRF power deposition. In particular, models indicate that finite orbit width effects lead to a reduction in the total stored energy and of the tail energy in the center of the plasma, relative to that predicted by the zero banana width models. In this paper, detailed comparisons between the calculated ICRF power deposition profiles and experimentally measured quantities will be presented which indicate that significant deviations from the zero banana width models occur even for modest power levels (P rf ∼ 6 MW) in the TFTR experiments

  1. Measurements of tritium recycling and isotope exchange in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kamperschroer, J.; Mueller, D.; Nagy, A.; Stotler, D.P.

    1996-05-01

    Tritium Balmer-alpha (T α ) emission, along with H α and D α is observed in the current D-T experimental campaign in TFTR. The data are a measure of the fueling of the plasma by tritium accumulated in the TFTR limiter and the spectral profile maps neutral hydrogenic velocities. T α is relatively slow to appear in tritium neutral beam heated discharges, (T α /(H α + D α + T α ) = 11% after 8 tritium-only neutral beam discharges). In contrast, the T α fraction in a sequence of six discharges fueled with tritium puff,s increased to 44%. Larger transient increases (up to 75% T α ) were observed during subsequent tritium gas puffs. Analysis of the Doppler broadened spectral profiles revealed overall agreement with the dissociation, charge exchange, sputtering and reflection velocities predicted by the neutral Monte-Carlo code DEGAS with some deficiency in the treatment of dissociation products in the 10--100 eV range

  2. Results obtained using the pellet charge exchange diagnostic on TFTR

    International Nuclear Information System (INIS)

    McChesney, J.M.; Fisher, R.K.; Parks, P.B.; Duong, H.H.; Mansfield, D.K.; Medley, S.S.; Roquemore, A.L.; Petrov, M.P.

    1994-05-01

    Experiments are underway on TFTR to measure the confined alpha particle distribution functions using small low-Z pellets injected into the plasma. Upon entering the plasma, the pellet ablates, forming a plasma ablation cloud, elongated in the magnetic field direction, that travels alongside the pellet. A small fraction of the fusion produced 3.5 MeV alpha particles incident on the cloud are converted to helium neutrals. By measuring the resultant helium neutrals escaping from the plasma by means of a mass and energy resolving charge exchange analyzer, the energy distribution of the alpha particles incident on the cloud can be inferred. Preliminary experiments to observe neutrals from the 100-1000 keV He tail produced during ICRF minority heating experiments were successful. However, no significant alpha particle signals have been observed during D-T operation on TFTR. The authors attribute this lack of signal to stochastic toroidal field ripple loss in the outer regions of the plasma. They are studying ways to improve the pellet penetration so that the pellet penetrates into the central regions of the plasma where ripple induced losses are small and the alpha population is high

  3. Configuration management of TFTR during final fabrication/assembly/installation

    International Nuclear Information System (INIS)

    Sabado, M.; Rappe, G.H.; Stern, E.; Wexler, H.

    1983-01-01

    In essence, configuration management consists of the establishment of a Baseline definition for each project phase, well documented, so that all project participants are conversant with it and the disciplined redefinition of the baseline as the project matures. This paper describes the methods by which the Baseline design for each phase of the TFTR program was updated. Definition was initiated through informal controls which became more formal as the design progressed. At the point where the design was essentially frozen, that is, released for procurement and manufacturing, a configuration change control procedure was instituted to continue on a routine basis both engineering and management review of all changes. Since the TFTR program is experimental in nature it was understood from the outset that desirable changes based on new analytical results and experimental results from other fusion programs could be injected into the design. The problem was one of maintaining the flexibility of providing a reasonable baseline definition, in order to allow the design to proceed yet avoiding the premature freezing of the design, in order to incorporate required changes at lowest cost

  4. Calibration issues of the TFTR multichannel neutron collimator

    International Nuclear Information System (INIS)

    Goeler, S. von; Johnson, L.C.; Bitter, M.; Efthimion, P.C.; Roquemore, A.L.

    1996-01-01

    The calibration procedures for the detectors in the Neutron Collimator are reviewed. The absolute calibration was performed for the NE451 detectors, in situ, by moving a DT neutron generator in the TFTR vacuum vessel across each sight line. This calibration was transferred to other detectors in the same channel. Four new sight lines have been installed at a different toroidal location, which view the plasma through the vacuum vessel port cover rather than through thinned windows. The new detectors are cross-calibrated to the NE451 detectors with a jog shot procedure, where the plasma is quickly shifted in major radius over a distance of 30 cm. The jog shot procedure shows that scattered neutrons account approximately for 30% of the signal of the new central channels. The neutron source strength from the collimator agrees within 10% with the source strength from global neutron monitors in the TFTR test cell. Detector non-linearity is discussed. Another special issue is the behavior of the detectors during T-puffs, where the DD/DT neutron ratio changes rapidly

  5. TFTR Inner Support Structure final assembly and installation

    International Nuclear Information System (INIS)

    Rocco, R.E.; Brown, G.; Carglia, G.; Heitzenroeder, P.; Koenig, F.; Mookerjee, S.; Raugh, J.

    1983-01-01

    The Inner Support Structure (ISS) of the TFTR provides a specific level of restraint to the net centering force and overturning moment produced by the Toroidal Field (TF) coils and to the vertical forces produced by the Inner Poloidal Field (PF) coils. This is accomplished consistent with the need for four radial dielectric breaks running the entire length of the ISS to prevent eddy current loops. A brief description of the major components, method of manufacture and material selection of the ISS and PF coils is presented. Particular attention is given to the integration of the PF coils and the ISS components into the total assembly and the installation of strain gauges and crack monitors on the ISS. The requirements of no gaps at the interfaces of the ISS teeth at all three horizontal planes is discussed. The problem encountered with achieving the no gap requirement and the successful resolution of this problem, including its impact on installation of the ISS, is also discussed. The installation of the ISS, including setting in position, preloading with TF coil clips, and final tensioning of the tension bars is discussed. A brief description of the lower and upper lead stem splicing operation is presented. Subsequent to the final assembly, electrical tests were performed prior to and after installation on the TFTR machine. An overview of the tests and their results is presented

  6. Computer code for the costing and sizing of TNS tokamaks

    International Nuclear Information System (INIS)

    Sink, D.A.; Iwinski, E.M.

    1977-01-01

    A FORTRAN code for the COsting And Sizing of Tokamaks (COAST) is described. The code was written to conduct detailed analyses on the engineering features of the next tokamak fusion device following TFTR. The ORNL/Westinghouse study of TNS (The Next Step) has involved the investigation of a number of device options, each over a wide range of plasma sizes. A generalized description of TNS is incorporated in the code and includes refined modeling of over forty systems and subsystems. Considerable detailed design and analyses have provided the basis for the thermal, electrical, mechanical, nuclear, chemical, vacuum, and facility engineering of the various subsystems. Currently, the code provides a tool for the systematic comparison of four toroidal field (TF) coil technologies allowing both D-shaped and circular coils. The coil technologies are: (1) copper (both room temperature and liquid-nitrogen cooled), (2) superconducting NbTi, (3) superconducting Nb 3 Sn, and (4) a Cu/NbTi/ hybrid. For the poloidal field (PF) coil systems copper conductors are assumed. The ohmic heating (OH) coils are located within the machine bore and have an air core, while the shaping field (SF) coils are located either within or outside the TF coils. The PF coil self and mutual inductances are calculated from the geometry, and the PF coil power supplies are modeled to account for time-dependent profiles for voltages and currents as governed by input data. Plasma heating is assumed to be by neutral beams, and impurity control is either passive or by a poloidal divertor system. The size modeling allows considerable freedom in specifying physics assumptions, operating scenarios, TF operating margin, and component geometric and performance parameters. Cost relationships have been developed for both plant and capital equipment and for annual utility and fuel expenses. The code has been used successfully to reproduce the sizing and costing of TFTR in order to calibrate the various models

  7. Tokamak devices: towards controlled fusion

    International Nuclear Information System (INIS)

    Trocheris, M.

    1975-01-01

    The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr

  8. What's happening at the edge of tokamaks

    International Nuclear Information System (INIS)

    Crandall, D.H.

    1987-01-01

    Handling the power deposition at the walls of a plasma fusion device and controlling the particle fueling of the plasma originated the interest in the edge of the plasma by magnetic fusion scientists. Recently this interest has intensified because of clear evidence that the quality of the central plasma confinement depends in unexpected ways on details of how the edge plasma is managed. Significant efforts are being pursued to understand and exploit the improved plasma confinement observed in the 'H-mode' obtained with divertors and in the 'super-shots' obtained with low neutral particle flux from the edge of TFTR limiter plasmas. The controls, that determine whether or not these well-confined plasmas are obtained, are applied in the edge plasma where a wealth of atomic and molecular processes occur. A qualitative overview of current research related to plasma edge and desirable features is presented to guide thoughts about atomic processes to be included in modeling and interpreting the plasma edge of tokamaks. (orig.)

  9. DEALS magnet concept and its applcations to high density, high field tokamak systems

    International Nuclear Information System (INIS)

    Hsieh, S.Y.; Powell, J.; Lehner, J.; Bezler, P.; Laverick, C.; Finkelman, M.; Brown, T.; Bundy, J.

    1977-01-01

    The goal of the DEALS program is to develop a demountable TF magnet system concept that will reduce construction and life cycle costs, enhance the accessibility of components inside the coil system, and increase the chances for being able to use large high-field magnet systems in post TFTR reactor experiments. These experiments are projected to occur during the mid 1980's, with conceptual designs beginning in two or three years. A number of recent studies have highlighted the need for Tokamak fusion reactor systems with reasonable down time for maintenance and repair and realistic operating capacity factors, as well as the need for smaller, lower cost reactors. Two scoping studies were carried out of recent Tokamak system concepts incorporating conventionally wound coils to assess the possibilities of using demountable coils of rectangular section with an active support system and a third more intensive study using a passive support with slight movement of the joints. These studies are described briefly

  10. Track-mounted remote handling system for the Tokamak Fusion Engineering Test

    International Nuclear Information System (INIS)

    Kelly, V.P.; Berger, J.D.; Daubert, R.L.; Yount, J.A.

    1982-01-01

    Concepts for remote handling machines (IVM) designed to transverse the interior of toroidal vessels with guidance and support from track systems have been developed for the proposed Tokamak Fusion Engineering Test (TFET). TFET has been proposed as an upgrade for the Tokamak Fusion Test Reactor (TFTR), currently nearing completion. The track-mounted IVMs were conceived to perform in-vessel remote maintenance for TFET, including removal and replacement of pump limiter blades and protective tiles as well as other maintenance-related tasks such as vessel wall inspection leak testing and interior cleanup. The conceptual IVMs consist of three manipulator arms mounted on a common frame member: a single power manipulator arm with high load carrying capacity and two lower-capacity servomanipulator arms. Descriptions of the IVM concepts, in-vessel track systems, and ex-vessel handling systems are presented

  11. Atomic physics studies of highly charged ions on tokamaks using x-ray spectroscopy

    International Nuclear Information System (INIS)

    Beiersdorfer, P.; von Goeler, S.; Bitter, M.; Hill, K.W.

    1989-07-01

    An overview is given of atomic physics issues which have been studied on tokamaks with the help resolution x-ray spectroscopy. The issues include the testing of model calculations predicting the excitation of line radiation, the determination of rate coefficients, and accurate atomic structure measurements. Recent research has focussed primarily on highly charged heliumlike (22 ≤ Z ≤ 28) and neonlike (34 ≤ Z ≤ 63) ions, and results are presented from measurements on the PLT and TFTR tokamaks. Many of the measurements have been aided by improved instrumental design and new measuring techniques. Remarkable agreement has been found between measurements and theory in most cases. However, in this review those areas are stressed where agreement is worst and where further investigations are needed. 19 refs., 13 figs., 2 tabs

  12. An amplitude and phase control system for the TFTR rf heating sources

    International Nuclear Information System (INIS)

    Cutsogeorge, G.

    1989-04-01

    Feedback loops that control the amplitude and phase of the rf heating sources on TFTR are described. The method for providing arc protection is also discussed. Block diagrams and Bode plots are included. 6 figs

  13. Tokamak Fusion Test Reactor D-T results

    International Nuclear Information System (INIS)

    Meade, D.M.

    1995-01-01

    Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, α confinement, α heating and possible α-driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of about 20MW of tritium and 14MW of deuterium neutral beams into the TFTR produced a plasma with a T-to-D density ratio of about 1 and yielding a maximum fusion power of about 9.2MW. The fusion power density in the core of the plasma was about 1.8MWm -3 , approximating that expected in a D-T fusion reactor. A TFTR plasma with a T-to-D density ratio of about 1 was found to have about 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass A of τ E ∝A 0.6 . The core ion temperature increased from 30 to 37keV owing to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 to 10.6keV can be attributed to electron heating by the α particles. The approximately 5% loss of α particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined high energy α particles and the resultant α ash density. At fusion power levels of 7.5MW, fluctuations at the toroidal Alfven eigen-mode frequency were observed by the fluctuation diagnostics. However, no additional α loss due to the fluctuations was observed. (orig.)

  14. Beta normal control of TFTR using fuzzy logic

    International Nuclear Information System (INIS)

    Lawson, J.E.; Bell, M.G.; Marsala, R.J.; Mueller, D.

    1995-01-01

    In TFTR plasmas heated by neutral beam injection, the fusion power yield increases rapidly with the plasma pressure. However, the pressure is limited by the onset of instabilities which may result in plasma disruptions that would have had an adverse effect on the performance of subsequent discharges and increase the risk of damage to internal components. The likelihood of disruption has been found to correlate with the normalized beta, defined as βN = 2 x 10 8 μ circle left angle p perpendicular to right angle a / BTIp where left angle p perpendicular to right angle is the volume-average plasma perpendicular pressure, a the mid-plane minor radius of the plasma, BT the toroidal magnetic field and Ip the plasma current. Other variables, such as the peaking of the plasma pressure and current profiles, have been found to influence the threshold of βN at which the probability of disruption begins to increase significantly. For TFTR plasmas with high fusion performance (TFTR ''supershots'') the probability of disruption has been found to increase rapidly for βN > 1.8. Since confinement in this regime is affected by plasma-wall interaction, which can vary from shot to shot, operation at high βN with preprogrammed heating power pulses can produce an unacceptably high risk of disruption. To reduce the risk of producing beta-limit disruptions during neutral beam heating experiments, a control system, the Neutral Beam Power Feedback System (NBPFS), has been developed to modulate the total heating power by switching individual neutral beam sources on and off in response to the evolution of the normalized beta so that the limit will not be exceeded. The value of βN is calculated in real time and transmitted to the NBPFS. The value of βN and its calculated time derivative are input to a fuzzy logic controller which implements a proportional-derivative control based on the difference between βN and a programmed reference level βNREF which can be programmed as a function

  15. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Weiyue; Du, Shijun

    2014-01-01

    Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

  16. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

    Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

  17. Advanced Tokamak Stability Theory

    Science.gov (United States)

    Zheng, Linjin

    2015-03-01

    The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.

  18. Neutronic design for the TFTR lithium blanket module

    International Nuclear Information System (INIS)

    Cheng, E.T.; Engholm, B.A.; Su, S.D.

    1981-01-01

    The preliminary design of a lithium blanket module (LBM) to be installed and tested in the TFTR has been performed under subcontract to PPPL and EPRI. The objectives of the LBM program are calculation and measurement of neutron fluences and tritium production in a breeding blanket module using state of art techniques, comparison of calculations with measurements, and acquisition of operational experience with a fusion reactor blanket module. The neutronic design of the LBM is one of the key areas of this program in which the LBM composition and geometry are optimized and the boundary material effects on the tritium production in the blanket module are explored. The concept of employing sintered Li/sub 2/O pellets in tubes is proposed for the blanket design

  19. TFTR horizontal high-resolution Bragg x-ray spectrometer

    International Nuclear Information System (INIS)

    Hill, K.W.; Bitter, M.; Tavernier, M.

    1984-11-01

    A bent quartz crystal spectrometer of the Johann type with a spectral resolution of lambda/Δlambda = 10,000 to 25,000 is used on TFTR to determine central plasma parameters from the spectra of heliumlike and lithiumlike metal impurity ions (Ti, Cr, Fe, and Ni). The spectra are observed along a central radial chord and are recorded by a position sensitive multiwire proportional counter with a spatial resolution of 250. Standard delay-line time-difference readout is employed. The data are histogrammed and stored in 64k of memory providing 128 time groups of 512-channel spectra. The central ion temperature and the toroidal plasma rotation are inferred from the Doppler broadening and Doppler shift of the K lines. The central electron temperature, the distribution of ionization states, and dielectronic recombination rates are obtained from satellite-to-resonance line ratios. The performance of the spectrometer is demonstrated by measurements of the Ti XXI K radiation

  20. Power loading on the first wall during disruptions in TFTR

    International Nuclear Information System (INIS)

    Janos, A.; Fredrickson, E.D.; McGuire, K.M.; Nagayama, Y.; Owens, D.K.; Wilfrid, E.

    1992-01-01

    Heating of the first wall of TFTR due to disruptions is investigated experimentally using an extensive array of thermocouples. By comparing results from discharges with and without disruptions, we extract effects due to the disruption alone. Disruptions preferentially heat the same areas which are heated during discharges without disruptions. Hot areas are inward protrusions or regions unshielded by neighboring areas. Peaking factors in the toroidal direction, defined as peak temperature divided by average toroidal temperature, as a function of poloidal angle, are calculated. For nondisruptive discharges, the peaking factor varies between 2 and 4. For the disruptive portion of a discharge only, the peaking factor near the midplane, where most of the energy is deposited, ranges from 3 to 5. Further away from the midplane, the peaking factor can reach 28, although the heat load is less in that region. (orig.)

  1. Measurements of edge density profile modifications during IBW on TFTR

    International Nuclear Information System (INIS)

    Hanson, G.R.; Bush, C.E.; Wilgen, J.B.

    1997-01-01

    Ion Bernstein wave (IBW) antennas are known to have substantial localized effects on the plasma edge. To allow better understanding and measurement of these effects, the TFTR edge reflectometer has been relocated to the new IBW antenna. This move was facilitated by the incorporation of a diagnostic access tube in the IBW antenna identical to the original diagnostic tube in the fast-wave (FW) antenna. This allowed the reflectometer launcher to simply be moved from the old FW antenna to the new IBW antenna. Only a moderate extension of the waveguide transmission line was required to reconnect the reflectometer to the launcher in its new location. Edge density profile modification during IBW experiments has been observed. Results from IBW experiments will be presented and contrasted to the edge density modifications previously observed during FW heating experiments

  2. TFTR Mirnov coil analysis at plasma start-up

    International Nuclear Information System (INIS)

    Harley, T.R.; Buchenauer, D.A.; Coonrod, J.; McGuire, K.M.

    1986-01-01

    The methods for finding poloidal and toroidal numbers of MHD oscillations from Mirnov coils are reviewed and modified. Examples of various MHD phenomena occurring during start-up on TFTR are illustrated. It is found that the MHD mode structure best fits a model with the toroidal correction included. A new algorithm which finds m,n numbers can accommodate toroidal effects which are manifested in the phase data. The algorithm can find m,n numbers with a given toroidal correction parameter lambda', (lambda' = 0 → cylindrical). This algorithm is also used to find the optimal value of lambda' automatically, eliminating the need for ''guesswork.'' The algorithm finds the best parameters to the fit much faster than more conventional computational techniques. 9 refs., 21 figs., 2 tabs

  3. Designs for a TFTR full-power pumped limiter

    International Nuclear Information System (INIS)

    Budny, R.

    1986-10-01

    A pumped-limiter system which would provide increased particle control and enhance the performance of full-power discharges is being considered for TFTR. The system consists of two toroidal belts located near the Zirconium-Aluminium (ZrAl) getter panels. The limiter blades would be made of carbon/carbon composite in order to have a very thin profile, allowing a large fraction of the scrape-off flux to be pumped. Simulations of the plasma scrape-off and neutral transport indicate that the limiter pumping should reduce the recycling coefficient by 10 to 25%. Simulations of central plasma processes indicate that the lowered recycling could increase Q/sub fusion/ by more than 100%. This paper discusses the designs and the performance predictions for the system

  4. Tokamak confinement scaling laws

    International Nuclear Information System (INIS)

    Connor, J.

    1998-01-01

    The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

  5. Tokamak concept innovations

    International Nuclear Information System (INIS)

    1986-04-01

    This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

  6. Operation of TFTR neutral beams with heavy ions

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Stevenson, T.N.; Wright, K.E.; Dudek, L.E.; Grisham, L.R.; Newman, R.A.; O'Connor, T.E.; Oldaker, M.E.; von Halle, A.; Williams, M.D.

    1991-07-01

    High Z neutral atoms have been injected into TFTR plasmas in an attempt to enhance plasma confinement through modification of the edge electric field. TFTR ion sources have extracted 9 A of 62 keV Ne + for up to 0.2 s during injection into deuterium plasmas, and for 0.5 s during conditioning pulses. Approximately 400 kW of Ne 0 have been injected from each of two ion sources. Operation was at full bending magnet current, with the Ne + barely contained on the ion dump. Beamline design modifications to permit operation up to 120 keV with krypton or xenon are described. Such ions are too massive to be deflected up to the ion dump. The plan, therefore, is to armor those components receiving these ions. Even with this armor, modest increases in the bending magnet current capability are necessary to safely reach 120 kV with Kr or Xe. Information relevant to heavy ion operation was also acquired when several ion sources were inadvertently operated with water contamination. Spectroscopic analysis of certain pathological pulses indicate that up to 6% of the extracted ions were water. After dissociation in the neutralizer, water yields oxygen ions which, as with Ne, Kr, and Xe, are under-deflected by the magnet. Damage to a calorimeter scraper, due to the focal properties of the magnet, has resulted. A magnified power density of 6 KW/cm 2 for 2 s, from ∼ 90 kW of O + , is the suspected cause. 11 refs., 4 figs

  7. TFTR ultrahigh-vacuum pumping system incorporating mercury diffusion pumps

    International Nuclear Information System (INIS)

    Sink, D.A.; Sniderman, M.

    1976-06-01

    The TFTR vacuum vessel will have a system of four 61 cm diameter mercury diffusion pumps to provide a base pressure in the 10 -8 to 10 -9 Torr range as well as a low impurity level within the vessel. The system, called the Torus Vacuum Pumping System (TVPS), will be employed with the aid of an occasional 250 0 C bakeout in situ as well as periodic applications of aggressive discharge cleaning. The TVPS is an ultrahigh-vacuum (UHV) system using no elastomers as well as being a closed system with respect to tritium or any tritiated gases. The backing system employing approximately 75 all-metal isolation valves is designed with the features of redundancy and flexibility employed in a variety of ways to meet the fundamental requirements and functions enumerated for the TVPS. Since the design, is one which is a modification of the conceptual design of the TVPS, those features which have changed are discussed. Calculations are presented for the major performance parameters anticipated for the TVPS and include conductances, effective pumping speeds, base pressures, operating parameters, getter pump parameters, and calculations of time constants associated with leak checking. Modifications in the vacuum pumping system for the guard regions on the twelve bellows sections are presented so that it is compatible with the main TVPS. The bellows pumping system consists of a mechanical pump unit, a zirconium aluminum getter pump unit and a residual gas analyzer. The control and management of the TVPS is described with particular attention given to providing both manual and automatic control at a local station and at the TFTR Central Control. Such operations as testing, maintenance, leak checking, startup, bakeout, and various other operations are considered in some detail. Various aspects related to normal pulsing, discharge cleaning, non-tritium operations and tritium operations are also taken into consideration. A cost estimate is presented

  8. Two-stream cyclotron radiative instabilities due to the marginally mirror-trapped fraction for fustion alphas in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Arunasalam, V.

    1995-07-01

    It is shown here that the marginally mirror-trapped fraction of the newly-born fusion alpha particles in the deuterium-tritium (DT) reaction dominated tokamak plasmas can induce a two-stream cyclotron radiative instability for the fast Alfven waves propagating near the harmonics of the alpha particle cyclotron frequency {omega}{sub c{alpha}}. This can explain both the experimentally observed time behavior and the spatially localized origin of the fusion product ion cyclotron emission (ICE) in TFTR at frequencies {omega} {approx} m{omega}{sub c{alpha}}.

  9. A New Interpretation of Alpha-particle-driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    R. Nazikian; G.J. Kramer; C.Z. Cheng; N.N. Gorelenkov; H.L. Berk; S.E. Sharapov

    2003-01-01

    The original description of alpha-particle-driven instabilities in the Tokamak Fusion Test Reactor (TFTR) in terms of Toroidal Alfvin Eigenmodes (TAEs) remained inconsistent with three fundamental characteristics of the observations: (i) the variation of the mode frequency with toroidal mode number, (ii) the chirping of the mode frequency for a given toroidal mode number, and (iii) the anti-ballooning density perturbation of the modes. It is now shown that these characteristics can be explained by observing that cylindrical-like modes can exist in the weak magnetic shear region of the plasma that then make a transition to TAEs as the central safety factor decreases in time

  10. Two-stream cyclotron radiative instabilities due to the marginally mirror-trapped fraction for fustion alphas in tokamaks

    International Nuclear Information System (INIS)

    Arunasalam, V.

    1995-07-01

    It is shown here that the marginally mirror-trapped fraction of the newly-born fusion alpha particles in the deuterium-tritium (DT) reaction dominated tokamak plasmas can induce a two-stream cyclotron radiative instability for the fast Alfven waves propagating near the harmonics of the alpha particle cyclotron frequency ω cα . This can explain both the experimentally observed time behavior and the spatially localized origin of the fusion product ion cyclotron emission (ICE) in TFTR at frequencies ω ∼ mω cα

  11. Alpha-Driven MHD and MHD-Induced Alpha Loss in TFTR DT Experiments

    Science.gov (United States)

    Chang, Zuoyang

    1996-11-01

    Theoretical calculation and numerical simulation indicate that there can be interesting interactions between alpha particles and MHD activity which can adversely affect the performance of a tokamak reactor (e.g., ITER). These interactions include alpha-driven MHD, like the toroidicity-induced-Alfven-eigenmode (TAE) and MHD induced alpha particle losses or redistribution. Both phenomena have been observed in recent TFTR DT experiments. Weak alpha-driven TAE activity was observed in a NBI-heated DT experiment characterized by high q0 ( >= 2) and low core magnetic shear. The TAE mode appears at ~30-100 ms after the neutral beam turning off approximately as predicted by theory. The mode has an amplitude measured by magnetic coils at the edge tildeB_p ~1 mG, frequency ~150-190 kHz and toroidal mode number ~2-3. It lasts only ~ 30-70 ms and has been seen only in DT discharges with fusion power level about 1.5-2.0 MW. Numerical calculation using NOVA-K code shows that this type of plasma has a big TAE gap. The calculated TAE frequency and mode number are close to the observation. (2) KBM-induced alpha particle loss^1. In some high-β, high fusion power DT experiments, enhanced alpha particle losses were observed to be correlated to the high frequency MHD modes with f ~100-200 kHz (the TAE frequency would be two-times higher) and n ~5-10. These modes are localized around the peak plasma pressure gradient and have ballooning characteristics. Alpha loss increases by 30-100% during the modes. Particle orbit simulations show the added loss results from wave-particle resonance. Linear instability analysis indicates that the plasma is unstable to the kinetic MHD ballooning modes (KBM) driven primarily by strong local pressure gradients. ----------------- ^1Z. Chang, et al, Phys. Rev. Lett. 76 (1996) 1071. In collaberation with R. Nazikian, G.-Y. Fu, S. Batha, R. Budny, L. Chen, D. Darrow, E. Fredrickson, R. Majeski, D. Mansfield, K. McGuire, G. Rewoldt, G. Taylor, R. White, K

  12. Tokamak control simulator

    International Nuclear Information System (INIS)

    Edelbaum, T.N.; Serben, S.; Var, R.E.

    1976-01-01

    A computer model of a tokamak experimental power reactor and its control system is being constructed. This simulator will allow the exploration of various open loop and closed loop strategies for reactor control. This paper provides a brief description of the simulator and some of the potential control problems associated with this class of tokamaks

  13. Finite pressure effects on the tokamak sawtooth crash

    International Nuclear Information System (INIS)

    Nishimura, Yasutaro

    1998-07-01

    The sawtooth crash is a hazardous, disruptive phenomenon that is observed in tokamaks whenever the safety factor at the magnetic axis is below unity. Recently, Tokamak Test Fusion Reactor (TFTR) experimental data has revealed interesting features of the dynamical pressure evolution during the crash phase. Motivated by the experimental results, this dissertation focuses on theoretical modeling of the finite pressure effects on the nonlinear stage of the sawtooth crash. The crash phase has been studied numerically employed a toroidal magnetohydrodynamic (MHD) initial value code deduced from the FAR code. For the first time, by starting from a concentric equilibrium, it has been shown that the evolution through an m/n = 1/1 magnetic island induces secondary high-n ballooning instabilities. The magnetic island evolution gives rise to convection of the pressure inside the inversion radius and builds up a steep pressure gradient across the island separatrix, or current sheet, and thereby triggers ballooning instabilities below the threshold for the axisymmetric equilibrium. Due to the onset of secondary ballooning modes, concomitant fine scale vortices and magnetic stochasticity are generated. These effects produce strong flows across the current sheet, and thereby significant modify the m = 1 driven magnetic reconnection process. The resultant interaction of the high-n ballooning modes with the magnetic reconnection process is discussed

  14. Kinetic theory of plasma adiabatic major radius compression in tokamaks

    International Nuclear Information System (INIS)

    Gorelenkova, M.V.; Gorelenkov, N.N.; Azizov, E.A.; Romannikov, A.N.; Herrmann, H.W.

    1998-01-01

    In order to understand the individual charged particle behavior as well as plasma macroparameters (temperature, density, etc.) during the adiabatic major radius compression (R-compression) in a tokamak, a kinetic approach is used. The perpendicular electric field from the Ohm close-quote s law at zero resistivity is made use of in order to describe particle motion during the R-compression. Expressions for both passing and trapped particle energy and pitch angle change are derived for a plasma with high aspect ratio and circular magnetic surfaces. The particle behavior near the passing trapped boundary during the compression is studied to simulate the compression-induced collisional losses of alpha particles. Qualitative agreement is obtained with the alphas loss measurements in deuterium-tritium (D-T) experiments in the Tokamak Fusion Test Reactor (TFTR) [World Survey of Activities in Controlled Fusion Research [Nucl. Fusion special supplement (1991)] (International Atomic Energy Agency, Vienna, 1991)]. The plasma macroparameters evolution at the R-compression is calculated by solving the gyroaveraged drift kinetic equation. copyright 1998 American Institute of Physics

  15. Electrical conductivity in tokamaks and extended neoclassical theory

    International Nuclear Information System (INIS)

    Segre, S.E.; Zanza, V.

    1992-01-01

    The electrical conductivity measurements reported from various tokamaks (D-III, PLT, TEXT, ASDEX, JT-60, TEXTOR, JET, TFTR) and compared with the usual neoclassical theory are here also compared with the extended neoclassical theory where the electron-electron collision rate is anomalous while the electron-ion collision rate remains Coulombian. It is found that, out of the 14 experiments considered, three are consistent with both the neoclassical and the extended neoclassical theories, four are consistent only with the extended neoclassical theory, and four are consistent with the neoclassical theory and also, within the experimental errors, not inconsistent with the extended neoclassical theory; the remaining three experiments appear to be incompatible with both theories. It is concluded that the extended neoclassical theory is in better agreement with conductivity experiments than the conventional neoclassical theory and, indeed, the extended theory is a serious candidate for explaining tokamak behaviour, since it accommodates naturally an anomalous electron thermal transport, which the conventional neoclassical theory is unable to do. (author). 31 refs, 1 fig

  16. Tokamaks. 2. ed.

    International Nuclear Information System (INIS)

    Wesson, John; Campbell, D.J.; Connor, J.W.

    1997-01-01

    It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)

  17. Neoclassical alpha-particle losses in tokamaks allowing for large orbit widths

    International Nuclear Information System (INIS)

    Cox, M.; O'Brien, M.R.; Zaitsev, F.S.

    1994-01-01

    Alpha-particle physics is of particular importance now that research into controlled fusion has reached thermonuclear parameters and D-T fuel has been used in JET and TFTR. Here we address the important topic of α-particle transport: if transport is too low helium ash accumulates quenching the burn; if it is too high heating of the plasma by fast α-particles is insufficient to maintain the burn. We give results from simulations of α-particle distributions (f α ) which self-consistently treat α-particle birth, collisional slowing down and neoclassical radial transport. The (steady-state) f α is calculated by the FPP code as a function of speed (v), pitch-angle (θ) and flux surface radius (r). This code is based on a 3D Fokker-Planck theory of 'banana regime' neoclassical effects in tokamaks which can treat large deviations of fast ion orbits from flux surfaces and non-Maxwellian distributions. The code reproduces standard neoclassical results for Maxwellian distributions in the large aspect ratio (ε) and small orbit width (Δ) limits (e.g. radial fluxes, conductivities and bootstrap currents), but can also be used for small ε and large Δ which are difficult to treat analytically. The code is particularly useful for α-particle studies as (a) the experimental evidence is that fast ion transport is usually consistent with neoclassical theory, unlike electron or thermal ion transport, and (b) trapped fast ion orbits can deviate greatly from flux surfaces. An alternative to this Fokker-Planck treatment is Monte Carlo modelling. However, representation of the detailed structure of f α (θ,v,r) would require very large number of particles, and hence be very slow. Calculations have been made for parameters typical of TFTR, JET, SSTR (an 'advanced tokamak' reactor) and STR (a tight aspect ratio or 'spherical' tokamak reactor, though only the JET results are discussed in detail. (author) 4 refs., 4 figs

  18. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, I., E-mail: yublinski@yandex.ru [FSUE ' Red Star' , Moscow (Russian Federation); Vertkov, A.; Evtikhin, V.; Balakirev, V.; Ionov, D.; Zharkov, M. [FSUE ' Red Star' , Moscow (Russian Federation); Tazhibayeva, I. [IAE NNC RK, Kurchatov (Kazakhstan); Mirnov, S. [TRINITI, Troitsk, Moscow Region (Russian Federation); Khomiakov, S.; Mitin, D. [OJSC Dollezhal Institute, Moscow (Russian Federation); Mazzitelli, G. [ENEA RC Frascati (Italy); Agostini, P. [ENEA RC Brasimone (Italy)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Black-Right-Pointing-Pointer Capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. Black-Right-Pointing-Pointer Lithium divertor module for KTM tokamak is under development. Black-Right-Pointing-Pointer Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Black-Right-Pointing-Pointer Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation. - Abstract: Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z{sub eff} of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of

  19. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  20. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

  1. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    Bickerton, R.J.

    1977-01-01

    The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)

  2. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

  3. TFTR neutral beam control and monitoring for DT operations

    International Nuclear Information System (INIS)

    O'Connor, T.; Kamperschroer, J.; Chu, J.

    1995-01-01

    Record fusion power output has recently been obtained in TFTR with the injection of deuterium and tritium neutral beams. This significant achievement was due in part to the controls, software, and data processing capabilities added to the neutral beam system for DT operations. Chief among these improvements was the addition of SUN workstations and large dynamic data storage to the existing Central Instrumentation Control and Data Acquisition (CICADA) system. Essentially instantaneous look back over the recent shot history has been provided for most beam waveforms and analysis results. Gas regulation controls allowing remote switchover between deuterium and tritium were also added. With these tools, comparison of the waveforms and data of deuterium and tritium for four test conditioning pulses quickly produced reliable tritium setpoints. Thereafter, all beam conditioning was performed with deuterium, thus saving the tritium supply for the important DT injection shots. The lookback capability also led to modifications of the gas system to improve reliability and to control ceramic valve leakage by backbiasing. Other features added to improve the reliability and availability of DT neutral beam operations included master beamline controls and displays, a beamline thermocouple interlock system, a peak thermocouple display, automatic gas inventory and cryo panel gas loading monitoring, beam notching controls, a display of beam/plasma interlocks, and a feedback system to control beam power based on plasma conditions

  4. Confinement studies of ohmically heated plasmas in TFTR

    International Nuclear Information System (INIS)

    Efthimion, P.C.; Bretz, N.L.; Bell, M.G.

    1985-03-01

    Systematic scans of density in large deuterium plasmas (a = 0.83 m) at several values of plasma current and toroidal magnetic field strength indicate that the total energy confinement time, tau/sub E/, is proportional to the line-average density anti n/sub e/ and the limiter q. Confinement times of approx. 0.3 s have been observed for anti n/sub e/ = 2.8 x 10 19 m -3 . Plasma size scaling experiments with plasmas of minor radii a = 0.83, 0.69, 0.55, and 0.41 m at constant limiter q reveal a confinement dependence on minor radius. The major-radius dependence of tau/sub E/, based on a comparison between TFTR and PLT results, is consistent with R 2 scaling. From the power balance, the thermal diffusivity chi/sub e/ is found to be significantly less than the INTOR value. In the a = 0.41 m plasmas, saturation of confinement is due to neoclassical ion conduction (chi/sub i/ neoclassical >> chi/sub e/)

  5. Operating experience with TFTR's Tritium Storage and Delivery System

    International Nuclear Information System (INIS)

    Voorhees, D.R.

    1995-01-01

    The Tritium Storage and Delivery System (TSDS) at TFTR was fabricated at Monsanto Mound Lab in the late 1970's and delivered to PPPL in the early 1980's. Commissioning progressed slowly and was finally completed in 1992 following a series of Preoperational tests and Integrated Systems tests. Those tests included thorough leak testing of glove boxes and process piping, electrical interlocks and controls, instrumentation calibrations, volume determinations and verification of uranium bed capacity. The system accepted tritium in dilute form in May of 1993 and began serious usage of pure tritium in November 1993. As the throughput of high purity tritium increased, shortcomings of the system became evident and extensive repairs were implemented. System leakage and material compatibility were the primary causes of the problems. To date, the system has received, stored and delivered over 500 kCi of tritium and is performing very well. The dedicated quadrupole mass spectrometer and beta scintillator system has been analyzing tritium bearing and pure gas streams for over 3 years with minimal downtime

  6. Cryogenic supplies for the TFTR neutral beam line cryopanels

    International Nuclear Information System (INIS)

    Pinter, G.

    1977-01-01

    Cryocondensing panels will be used for the Neutral Beam Lines of the TFTR to satisfy a pumping speed requirement of 2.5 x 10 6 l/s. The cryocondensing panels are fed by liquid helium (LHe), boiling at selectable temperatures of 4.5 0 K or 3.8 0 K. Liquid nitrogen (LN 2 ) panels and chevrons thermally shield the LHe panel. The closed-loop LHe supply system and the open loop LN 2 system are discussed. The helium refrigerator of minimum 1070-W capacity, together with its distribution system, and the nitrogen distribution system in the ton/hour LN 2 range is presented. Problems and their solutions in connection with the LHe system, including the distribution over a distance of 500 feet of large quantities of liquid/gas mixtures with load variations over the range of about 3 : 1, and the economies of various types of distribution lines (passive, pumped, shielded, combined), are described. The system design passed the preliminary phase. Design features and auxiliary equipment to assure dispersion of large quantities of nitrogen into the atmosphere and to permit operation under degraded cryogenic helium refrigerator performance are also discussed in Design Considerations

  7. TFTR 60 GHz alpha particle collective Thomson Scattering diagnostic

    International Nuclear Information System (INIS)

    Machuzak, J.S.; Woskov, P.P.; Gilmore, J.; Bretz, N.L.; Park, H.K.; Bindslev, H.

    1995-03-01

    A 60 GHz gyrotron collective Thomson Scattering alpha particle diagnostic has been implemented for the D-T period on TFM. Gyrotron power of 0.1-1 kW in pulses of up to 1 second can be launched in X-mode. Efficient corrugated waveguides are used with antennaes and vacuum windows of the TFTR Microwave Scattering system. A multichannel synchronous detector receiver system and spectrum analyzer acquire the scattered signals. A 200 Megasample/sec digitizer is used to resolve fine structure in the frequency spectrum. By scattering nearly perpendicular to the magnetic field, this experiment will take advantage of an enhancement of the scattered signal which results from the interaction of the alpha particles with plasma resonances in the lower hybrid frequency range. Significant enhancements are expected, which will make these measurements possible with gyrotron power less than 1 kW, while maintaining an acceptable signal to noise ratio. We hope to extract alpha particle density and velocity distribution functions from the data. The D and T fuel densities and temperatures may also be obtainable by measurement of the respective ion cyclotron harmonic frequencies

  8. Explosion potential of neutral-beam source cryopumps for TFTR

    International Nuclear Information System (INIS)

    Graham, W.G.; Lim, T.H.; Ruby, L.

    1977-12-01

    The explosion potential of the test cryopump became a paramount issue in the safety analysis required for the reactor experiment. The administrative limit for loading of the cryopump with normal hydrogen or deuterium is that amount of gas which will produce a partial pressure of 13 torr at a total pressure of 1 atmosphere, i.e., a 1.7% mixture by volume. At atmospheric pressure, combustion can occur for mixtures in the range 4.0 to 75%. It is important to know whether, in a leak-up-to-air accident, when the partial pressure will range from 100% to 1.7%, an explosion can occur. For the test cryopump (250l), loaded to the administrative limit, the energy of combustion would amount to 9.21 x 10 5 J, or 21.9 g of T.N.T. equivalent. However, for a TFTR beamline (73,000l), the corresponding numbers are 2.69 x 10 7 J, or 6.39 x 10 3 g of T.N.T. equivalent

  9. NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0.8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Dickens, J.K.; Hill, N.W.; Hou, F.S.; McConnell, J.W.; Spencer, R.R.; Tsang, F.Y.

    1985-08-01

    A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report

  10. Pitch angle resolved measurements of escaping charged fusion products in TFTR

    International Nuclear Information System (INIS)

    Zweben, S.J.

    1989-01-01

    Measurements of the flux of charged fusion products escaping from the TFTR plasma have been made with a new type of detector which can resolve the particle flux vs. pitch angle, energy, and time. The design of this detector is described, and results from the 1987 TFTR run are presented. These results are roughly consistent with predictions from a simple first-orbit particle loss model with respect to the pitch angle, energy, time, and plasma current dependence of the signals. 11 refs., 9 figs

  11. Pitch angle resolved measurements of escaping charged fusion products in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Zweben, S.J.

    1989-01-01

    Measurements of the flux of charged fusion products escaping from the TFTR plasma have been made with a new type of detector which can resolve the particle flux vs. pitch angle, energy, and time. The design of this detector is described, and results from the 1987 TFTR run are presented. These results are roughly consistent with predictions from a simple first-orbit particle loss model with respect to the pitch angle, energy, time, and plasma current dependence of the signals. 11 refs., 9 figs.

  12. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

  13. Gas utilization in the Tokamak Fusion Test Reactor neutral beam injectors

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Gammel, G.M.; Kugel, H.W.; Grisham, L.R.; Stevenson, T.N.; von Halle, A.; Williams, M.D.; Jones, T.T.C.

    1989-01-01

    Measurements of gas utilization were performed using hydrogen and deuterium beams in the Tokamak Fusion Test Reactor (TFTR) neutral beam test beamline to study the feasibility of operating tritium beams with existing ion sources under conditions of minimal tritium consumption. (i) It was found that the fraction of gas molecules introduced into the TFTR long-pulse ion sources that are converted to extracted ions (i.e., the ion source gas efficiency) was higher than with previous short-pulse sources. Gas efficiencies were studied over the range 33%--55%, and its effect on neutralization of the extracted ions was studied. At the high end of the gas efficiency range, the neutral fraction of the beam fell below that predicted from room-temperature molecular gas flow (similar to observations at the Joint European Torus). (ii) Beam isotope change studies were performed. No extracted hydrogen ions were observed in the first deuterium beam following a working gas change from H 2 to D 2 . There was no arc conditioning or gas injection preceding the first beam extraction attempt. (iii) Experiments were also performed to determine the reliability of ion source operation during the long waiting periods between pulses that are anticipated during tritium operation. It was found that an ion source conditioned to 120 kV could produce a clean beam pulse after a waiting period of 14 h by preceding the beam extraction with several acceleration voltage/filament warm-up pulses. It can be concluded that the operation of up to six ion sources on tritium gas should be compatible with on-site inventory restrictions established for D--T, Q = 1 experiments on TFTR

  14. Measurements of charged fusion product diffusion in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Boivin, R.L.

    1991-12-01

    The single particle confinement of charged fusion products, namely the 1 MeV triton and the 3 MeV proton, has been studied using a detector located near the outer midplane of TFTR. The detector, which measure the flux of escaping particles, is composed of a scintillator (ZnS(Ag)) and a system of collimating apertures, which permit pitch angle, energy and time resolution. It is mounted on a movable probe which can be inserted 25 cm into the vacuum vessel. Measurements indicate a level of losses higher than expected from a first-orbit loss mechanism alone. The primary candidate for explaining the observed anomalous losses is the toroidal field (TF) stochastic ripple diffusion, theoretically discovered by Goldston, White and Boozer. This loss mechanism is expected to be localized near the outer midplane where, at least at high current ({approx gt} 1.0 MA) it would locally dominate over first-orbit losses. Calculations made with a mapping particle orbit code (MAPLOS) show a semi-quantitative agreement with the measurements. The predominant uncertainties in the numerical simulations were found to originate from the modeling of the first wall geometry and also from the assumed plasma current and source profiles. Direct measurements of the diffusion rate were performed by shadowing the detector with a second movable probe used as an obstacle. The diffusion rate was also measured by moving the detector behind the radius of the RF limiters, located on the outer wall. Comparisons of these experimental results with numerical simulations, which include diffusive mechanisms, indicate a quantitative agreement with the TF stochastic ripple diffusion model.

  15. Measurements of charged fusion product diffusion in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Boivin, Rejean Louis [Princeton Univ., NJ (United States)

    1991-12-01

    The single particle confinement of charged fusion products, namely the 1 MeV triton and the 3 MeV proton, has been studied using a detector located near the outer midplane of TFTR. The detector, which measure the flux of escaping particles, is composed of a scintillator [ZnS(Ag)] and a system of collimating apertures, which permit pitch angle, energy and time resolution. It is mounted on a movable probe which can be inserted 25 cm into the vacuum vessel. Measurements indicate a level of losses higher than expected from a first-orbit loss mechanism alone. The primary candidate for explaining the observed anomalous losses is the toroidal field (TF) stochastic ripple diffusion, theoretically discovered by Goldston, White and Boozer. This loss mechanism is expected to be localized near the outer midplane where, at least at high current (≳ 1.0 MA) it would locally dominate over first-orbit losses. Calculations made with a mapping particle orbit code (MAPLOS) show a semi-quantitative agreement with the measurements. The predominant uncertainties in the numerical simulations were found to originate from the modeling of the first wall geometry and also from the assumed plasma current and source profiles. Direct measurements of the diffusion rate were performed by shadowing the detector with a second movable probe used as an obstacle. The diffusion rate was also measured by moving the detector behind the radius of the RF limiters, located on the outer wall. Comparisons of these experimental results with numerical simulations, which include diffusive mechanisms, indicate a quantitative agreement with the TF stochastic ripple diffusion model.

  16. Measurements of charged fusion product diffusion in TFTR

    International Nuclear Information System (INIS)

    Boivin, R.L.

    1991-12-01

    The single particle confinement of charged fusion products, namely the 1 MeV triton and the 3 MeV proton, has been studied using a detector located near the outer midplane of TFTR. The detector, which measure the flux of escaping particles, is composed of a scintillator [ZnS(Ag)] and a system of collimating apertures, which permit pitch angle, energy and time resolution. It is mounted on a movable probe which can be inserted 25 cm into the vacuum vessel. Measurements indicate a level of losses higher than expected from a first-orbit loss mechanism alone. The primary candidate for explaining the observed anomalous losses is the toroidal field (TF) stochastic ripple diffusion, theoretically discovered by Goldston, White and Boozer. This loss mechanism is expected to be localized near the outer midplane where, at least at high current (approx-gt 1.0 MA) it would locally dominate over first-orbit losses. Calculations made with a mapping particle orbit code (MAPLOS) show a semi-quantitative agreement with the measurements. The predominant uncertainties in the numerical simulations were found to originate from the modeling of the first wall geometry and also from the assumed plasma current and source profiles. Direct measurements of the diffusion rate were performed by shadowing the detector with a second movable probe used as an obstacle. The diffusion rate was also measured by moving the detector behind the radius of the RF limiters, located on the outer wall. Comparisons of these experimental results with numerical simulations, which include diffusive mechanisms, indicate a quantitative agreement with the TF stochastic ripple diffusion model

  17. Comparison of explicit calculations for n = 3 to 8 dielectronic satellites of the FeXXV Kα resonance line with experimental data from the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Decaux, V.; Bitter, M.; Hsuan, H.; Hill, K.W.; von Goeler, S.; Park, H.; Bhalla, C.P.

    1991-12-01

    Dielectronic satellite spectra of the FeXXV Kα resonance line observed from the Tokamak Fusion Test Reactor (TFTR) plasmas have been compared with recent explicit calculations for the n = 3 to 8 dielectronic satellites as well as the earlier theoretical predictions, which were based on the 1/n 3 scaling law for n > 4 satellites. The analysis has been performed by least-squares fits of synthetic spectra to the experimental data. The synthetic spectra constructed from both theories are in good agreement with the observed data. However, the electron temperature values obtained from the fit of the present explicit calculations are in better agreement with independent measurements. 20 refs., 4 figs

  18. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

  19. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Bosco, E. Del; Malaquias, A.; Mank, G.; Oost, G. van; He, Yexi; Hegazy, H.; Hirose, A.; Hron, M.; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

  20. Studies of the permeation and diffusion of tritium and hydrogen in TFTR

    International Nuclear Information System (INIS)

    Garber, H.J.

    1975-10-01

    This report documents the main features of studies conducted on the permeation and diffusion of tritium and hydrogen through the stainless steel sections comprising the vacuum vessel of TFTR. The overall conclusion of these studies is that tritium releases to the environment resulting from TFTR operations under normal conditions will be very small, less than one curie per year. A basis is described for predicting the magnitudes of the applicable transport properties for tritium-austenitic stainless steel systems as derived from a survey of the technical literature on tritium transport. The key characteristics of the TFTR vacuum vessel that are involved in the permeation and diffusion calculations are given. Information is given regarding the contemplated plasma scenarios and associated required gas injection quantities. Various issues involved in the bakeout of the vacuum vessel are discussed; focussing principally on the problems associated with in-situ bakeout and related means to reduce outgassing from the TFTR vessel and vacuum pumping system hardware. The anticipated tritium releases are studied considering the diffusion transients

  1. Facility for the testing of the TFTR prototype neutral beam injector

    Energy Technology Data Exchange (ETDEWEB)

    Haughian, J.M.

    1977-07-01

    The design of the prototype neutral beam injection system for TFTR is nearing completion at the Lawrence Livermore Laboratory. This paper describes some of the features of the facility at the Lawrence Berkeley Laboratory where this prototype will be assembled and tested.

  2. Facility for the testing of the TFTR prototype neutral beam injector

    International Nuclear Information System (INIS)

    Haughian, J.M.

    1977-07-01

    The design of the prototype neutral beam injection system for TFTR is nearing completion at the Lawrence Livermore Laboratory. This paper describes some of the features of the facility at the Lawrence Berkeley Laboratory where this prototype will be assembled and tested

  3. Transport simulations TFTR: Theoretically-based transport models and current scaling

    International Nuclear Information System (INIS)

    Redi, M.H.; Cummings, J.C.; Bush, C.E.; Fredrickson, E.; Grek, B.; Hahm, T.S.; Hill, K.W.; Johnson, D.W.; Mansfield, D.K.; Park, H.; Scott, S.D.; Stratton, B.C.; Synakowski, E.J.; Tang, W.M.; Taylor, G.

    1991-12-01

    In order to study the microscopic physics underlying observed L-mode current scaling, 1-1/2-d BALDUR has been used to simulate density and temperature profiles for high and low current, neutral beam heated discharges on TFTR with several semi-empirical, theoretically-based models previously compared for TFTR, including several versions of trapped electron drift wave driven transport. Experiments at TFTR, JET and D3-D show that I p scaling of τ E does not arise from edge modes as previously thought, and is most likely to arise from nonlocal processes or from the I p -dependence of local plasma core transport. Consistent with this, it is found that strong current scaling does not arise from any of several edge models of resistive ballooning. Simulations with the profile consistent drift wave model and with a new model for toroidal collisionless trapped electron mode core transport in a multimode formalism, lead to strong current scaling of τ E for the L-mode cases on TFTR. None of the theoretically-based models succeeded in simulating the measured temperature and density profiles for both high and low current experiments

  4. Feasibility of laser pumping with neutron fluxes from present-day large tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L.

    1986-08-01

    The minimum fusion-neutron flux needed to observe nuclear-pumped lasing with tokamaks can be reduced substantially by optimizing neutron scattering into the laser cell, located between adjacent toroidal-field coils. The laser lines most readily pumped are probably the /sup 3/He-Ne lines at 0.633 ..mu.. and in the infrared, where the /sup 3/He-Ne gas is excited by energetic ions produced in the /sup 3/He(n,p)T reaction. These lines are expected to lase at the levels of D-T neutron flux foreseen for the TFTR in 1989 (>>10/sup 12/ n/cm/sup 2//s), while amplification should be observable at the existing levels of D-D neutron flux (greater than or equal to 5 x 10/sup 9/ n/cm/sup 2//s). Lasing on the 1.73 ..mu.. and 2.63 ..mu.. transitions of Xe may be observable at the maximum expected levels of D-T neutron flux in TFTR enhanced by scattering.

  5. Feasibility of laser pumping with neutron fluxes from present-day large tokamaks

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-08-01

    The minimum fusion-neutron flux needed to observe nuclear-pumped lasing with tokamaks can be reduced substantially by optimizing neutron scattering into the laser cell, located between adjacent toroidal-field coils. The laser lines most readily pumped are probably the 3 He-Ne lines at 0.633 μ and in the infrared, where the 3 He-Ne gas is excited by energetic ions produced in the 3 He(n,p)T reaction. These lines are expected to lase at the levels of D-T neutron flux foreseen for the TFTR in 1989 (>>10 12 n/cm 2 /s), while amplification should be observable at the existing levels of D-D neutron flux (≥ 5 x 10 9 n/cm 2 /s). Lasing on the 1.73 μ and 2.63 μ transitions of Xe may be observable at the maximum expected levels of D-T neutron flux in TFTR enhanced by scattering

  6. Advanced commercial tokamak study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

    1985-12-01

    Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs

  7. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-03-01

    This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

  8. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

  9. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1993-04-01

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

  10. Magnetic ''islandography'' in tokamaks

    International Nuclear Information System (INIS)

    Callen, J.D.; Waddell, B.V.; Hicks, H.R.

    1978-09-01

    Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks

  11. Accelerator technology in tokamaks

    International Nuclear Information System (INIS)

    Kustom, R.L.

    1977-01-01

    This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

  12. Nonlinear ω*-stabilization of the m = 1 mode in tokamaks

    International Nuclear Information System (INIS)

    Rogers, B.; Zakharov, L.

    1995-08-01

    Earlier studies of sawtooth oscillations in Tokamak Fusion Test Reactor supershots (Levinton et al, Phys. Rev. Lett. 72, 2895 (1994); Zakharov, et al, Plasma Phys. and Contr. Nucl. Fus. Res., Proc. 15th Int. Conf., Seville 1994, Vienna) have found an apparent contradiction between conventional linear theory and experiment: even in sawtooth-free discharges, the theory typically predicts instability due to a nearly ideal m = 1 mode. Here, the nonlinear evolution of such mode is analyzed using numerical simulations of a two-fluid magnetohydrodynamic (MHD) model. We find the mode saturates nonlinearly at a small amplitude provided the ion and electron drift-frequencies ω* i,e are somewhat above the linear stability threshold of the collisionless m = 1 reconnecting mode. The comparison of the simulation results to m = 1 mode activity in TFTR suggests additional, stabilizing effects outside the present model are also important

  13. Electron cyclotron measurements with the fast scanning heterdyne radiometer on the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Taylor, G.; Efthimion, P.C.; McCarthy, M.P.; Fredd, E.A.; Cutler, R.C.

    1986-01-01

    Three fast scanning heterodyne receivers, swept between 75-110 GHz, 110-170 GHz, and 170-210 GHz, have measured electron cyclotron emission on the horizontal midplane of the tokamak fusion test reactor (TFTR) plasma. A second harmonic microwave mixer in the 170-210 GHz receiver allows the use of a 75-110 GHz backward wave oscillator as a swept local oscillator. Electron temperature profile evolution data with a time resolution of 2 msec and a profile acquisition rate of 250 Hz are presented for gas-fuelled and pellet-fuelled ohmic and neutral beam heated plasmas with toroidal fields up to 5.2 tesla. Recent results from a swept mode absolute calibration technique which can improve the accuracy and data collection efficiency during in-situ calibration are also presented

  14. ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

  15. Internal disruption in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1990-01-01

    A review of results of experimental and theoretical investigations of internal disruption in tokamaks is given. Specific features of various types of saw-tooth oscillations are described and their classification is performed. Theoretical models of the process of development of internal disruption instability are discussed. Effect of internal disruption on parameters of plasma, confined in tokamak, is considered. Scalings of period and amplitude of saw-tooth oscillations, as well as version radius are presented. Different methods for stabilizing instability of internal disruption are described

  16. Overview of Tokamak Results

    International Nuclear Information System (INIS)

    Unterberg, Bernhard; Samm, Ulrich

    2004-01-01

    An overview is given of recent results obtained in tokamak devices. We introduce basic confinement scenarios as L-mode, H-mode and plasmas with an internal transport barrier and discuss methods for profile control. Important findings in DT-experiments at JET as α-particle heating are described. Methods for power exhaust like plasma regimes with a radiating mantle and radiative divertor scenarios are discussed. The overall impact of plasma edge conditions on the general plasma performance in tokamaks is illustrated by describing the impact of wall conditions on confinement and the edge operational diagram of H-mode plasmas

  17. Internal disruptions in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1990-01-01

    Experimental and theoretical studies of the phenomenon of internal disruptions in tokamaks are reviewed. A classification scheme is introduced and the features of different types of sawtooth oscillations are described. A theoretical model for the development of the internal disruption instability is discussed. The effect of internal disruptions on the parameters of plasma confined in tokamaks is discussed. Scaling laws for the period and amplitude of sawtooth oscillations, as well as for the inversion radius, are presented. Different methods of stabilizing the internal disruption instability are described

  18. Measurement of ion profiles in TFTR neutral beamlines

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Grisham, L.R.; Kugel, H.W.; O'Connor, T.E.; Stevenson, T.N.; von Halle, A.; Williams, M.D.

    1992-02-01

    A technique is described whereby the ion dumps inside the TFTR Neutral Beam Test Stand were used to measure thermal profiles of the full-, half-, and third-energy ions. 136 thermocouples were installed on the full-energy ion dump, allowing full beam contours. Additional linear arrays across the widths of the half- and third-energy ion dumps provided a measure of the shape, in the direction parallel to the grid rails, of the half- and third-energy ions, and, hence, of the molecular ions extracted from the source. As a result of these measurements it was found that the magnet was more weakly focusing, by a factor of two, than expected, explaining past overheating of the full-energy ion dump. Hollow profiles on the half- and third-energy ion dumps were observed, suggesting that extraction of D 2 + and D 3 + are primarily from the edge of the ion source. If extraction of half-energy ions is from the edge of the accelerator, a divergence parallel to the grid rails of 0.6 degrees ±0.1 degrees results. It is postulated that a nonuniform gas profile near the accelerator is the cause of the hollow partial-energy ion profiles; the pressure being depressed over the accelerator by particles passing through this highly transparent structure. Primary electrons reaching the accelerator produce nonuniform densities of D 2 + through the ionization of this across the full-energy dump was examined as a means of reducing the power density. By unbalancing the current in the two coils of the magnet, on a shot by shot basis, by up to 2:1 ratio, it was possible to move the centerline of the full-energy ion beam sideways by ∼12.5 cm. The adoption of such a technique, with a ramp of the coil imbalance from 2:1 to 1:2 over a beam pulse, could reduce the power density by a factor of ≥1.5

  19. Ideal MHD stability of high poloidal beta equilibria in TFTR

    International Nuclear Information System (INIS)

    Sabbagh, S.A.; Mauel, M.E.; Navratil, G.A.; Bell, M.G.; Budny, R.V.; Chance, M.S.; Fredrickson, E.D.; Jardin, S.C.; Manickam, J.; McCune, D.C.; McGuire, K.M.; Wieland, R.M.; Zarnstorff, M.C.; Phillips, M.W.; Hughes, M.H.; Kesner, J.

    1991-01-01

    Recent experiments in TFTR have expanded the operating space of the device to include plasmas with values of var-epsilon β p dia ≡ 2μ 0 var-epsilon perpendicular >/ p >> 2 as large as 1.6, and Troyon normalized diamagnetic beta β N dia ≡ β t perpendicular aB t /10 -8 I p as large as 4.7. At values of var-epsilon β p dia ≥ 1.3, a separatrix was observed to enter the vacuum vessel, producing a naturally diverted discharge. Plasmas with large values of var-epsilon β p dia were created with both the plasma current, I p , held constant and with I p decreased, or ramped down, before the start of neutral beam injection. A convenient characterization of the change in I p using experimental parameters can be defined by the ratio of I p before the ramp down, to I p during the neutral beam heating phase, F I p . The ideal MHD stability of these equilibria is investigated to determine their location in stability space, and to study the role of plasma current and pressure profile modification in the creation of these high var-epsilon β p and β N plasmas. The evolution of these plasmas is modelled from experimental data using the TRANSP code. Two-dimensional equilibria are computed from the TRANSP results and used as input to both high and low-n stability codes including PEST. The high var-epsilon β p equilibria, which generally have an oblate cross-sectional shape, are in the first stability region to high-n ballooning modes. At constant I p , these equilibria generally have maximum pressure gradients near the magnetic axis and are stable to n=1 modes without a stabilizing conducting wall. The effect of the current profile shape on the stability of low-n kink/ballooning modes and the requirements for these plasmas to access the second stability region are examined. 6 refs

  20. High beta tokamaks

    International Nuclear Information System (INIS)

    Dory, R.A.; Berger, D.P.; Charlton, L.A.; Hogan, J.T.; Munro, J.K.; Nelson, D.B.; Peng, Y.K.M.; Sigmar, D.J.; Strickler, D.J.

    1978-01-01

    MHD equilibrium, stability, and transport calculations are made to study the accessibility and behavior of ''high beta'' tokamak plasmas in the range β approximately 5 to 15 percent. For next generation devices, beta values of at least 8 percent appear to be accessible and stable if there is a conducting surface nearby

  1. Sawtooth phenomena in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1989-01-01

    A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs

  2. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    Pare, V.K.

    1983-01-01

    Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging

  3. Research using small tokamaks

    International Nuclear Information System (INIS)

    1993-01-01

    This document consists of a collection of papers presented at the IAEA Technical Committee Meeting on Research Using Small Tokamaks. It contains 22 papers on a wide variety of research aspects, including diagnostics, design, transport, equilibrium, stability, and confinement. Some of these papers are devoted to other concepts (stellarators, compact tori). Refs, figs and tabs

  4. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

    The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

  5. Compact tokamak reactors

    International Nuclear Information System (INIS)

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1997-01-01

    The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. 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 tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, 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 recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs

  6. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1990-04-01

    This paper discusses the following work on the text tokamak: data systems; particle confinement; impurity transport; plasma rotation; runaway electrons; electron cyclotron heating; FIR system; transient transport; internal turbulence; edge turbulence; ion temperature; EML experiments; impurity pellet experiments; MHD experiments and analysis; TEXT Upgrade; and Upgrade diagnostics

  7. Structural analysis of TFTR toroidal field coil conceptual design

    International Nuclear Information System (INIS)

    Smith, R.A.

    1975-10-01

    The conceptual design evaluation of the V-shaped toroidal field coils on the Tokamak Fusion Test Reactor has been performed by detailed structural analysis with the finite element method. The innovation provided by this design and verified in this work is the capability to support toroidal field loads while simultaneously performing the function of twist restraint against the device axial torques resulting from the vertical field loads. The evaluations made for the conceptual design provide predictions for coil deflections and stresses. The results are available for the separate effects from toroidal fields, poloidal fields, and the thermal expansion of the coils as well as for the superposition of the primary loads and the primary plus thermal loads

  8. Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Hill, K.W.; Strachan, J.D.; Bell, M.G.; Scott, S.D.; Budny, R.; Marmar, E.S.; Snipes, J.A.; Terry, J.L.; Batha, S.; Bell, R.E.; Bitter, M.; Bush, C.E.; Chang, Z.; Darrow, D.S.; Ernst, D.; Fredrickson, E.; Grek, B.; Herrmann, H.W.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Levinton, F.M.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.; Ramsey, A.T.; Roquemore, A.L.; Skinner, C.H.; Stevenson, T.; Stratton, B.C.; Synakowski, E.; Taylor, G.; von Halle, A.; von Goeler, S.; Wong, K.L.; Zweben, S.J.

    1996-01-01

    Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium endash tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 10 21 m -3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral-beam heating is begun. copyright 1996 American Institute of Physics

  9. Tritium Removal by Laser Heating and Its Application to Tokamaks

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Guttadora, G.; Carpe, A.; Langish, S.; Young, K.M.; Nishi, M.; Shu, W.

    2001-01-01

    A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm 2 , and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed

  10. Analysis of IBW-driven plasma flows in tokamaks

    International Nuclear Information System (INIS)

    Berry, L.A.; Jaeger, E.F.; D'Azevedo, E.F.; Batchelor, D.B.; Carlsson, J.A.; Carter, M.D.; Cesario, R.

    2001-01-01

    Both theory and experiment have suggested that damping of Ion Bernstein Waves (IBWs) at ion cyclotron frequency harmonics could drive poloidal flows and lead to enhanced confinement for tokamaks. However, the early analyses were based on Reynolds stress closures of moment equations. More rigorous, finite Larmor radius (FLR) expansions of the radio frequency (RF) kinetic pressure for low harmonic interactions indicated that the Reynolds stress approximation was not generally valid, and resulted in significant changes in the plasma flow response. These changes were largest for wave interactions driven by finite Larmour radius effects. To provide a better assessment of higher harmonic interactions and IBW flow drive prospects, the electromagnetic (E and M) and RF kinetic force models are extended with no assumptions regarding the smallness of the ion Larmor radius. For both models, a spectral-width approximation was used to make the numerical analysis tractable. In addition, it was necessary to include the effects of plasma equilibrium gradients on the plasma conductivity and the RF-induced momentum in order to conserve energy and momentum. The analysis of high-harmonic IBW interactions for TFTR and FTU parameters indicates significant poloidal flow shears (relative to turbulence correlation times) for power levels available in present experiments. Recent advances in all-orders calculations of E and M fields in 2-D are also discussed. (author)

  11. Spectra of heliumlike krypton from tokamak fusion test reactor plasmas

    International Nuclear Information System (INIS)

    Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M.; Smith, A.; Fraenkel, B.

    1993-04-01

    Krypton has been injected into ohmically-heated TFTR plasmas with peak electron temperatures of 6 key to study the effects of krypton on the plasma performance and to investigate the emitted krypton line radiation, which is of interest for future-generation tokamaks such as ITER, both as a diagnostic of the central ion temperature and for the control of energy release from the plasma by radiative cooling. The emitted radiation was monitored with a bolometer array, an X-ray pulse height analysis system, and a high-resolution Johann-type crystal spectrometer; and it was found to depend very sensitively on the electron temperature profile. Satellite spectra of heliumlike krypton, KrXXXV, near 0.95 Angstrom including lithiumlike, berylliumlike and boronlike features were recorded in second order Bragg reflection. Radiative cooling and reduced particle recycling at the plasma edge region were observed as a result of the krypton injection for all investigated discharges. The observations are in reasonable agreement with modeling calculations of the krypton ion charge state distribution including radial transport

  12. Operation and control of high density tokamak reactors

    International Nuclear Information System (INIS)

    Attenberger, S.E.; McAlees, D.G.

    1976-01-01

    The incentive for high density operation of a tokamak reactor was discussed. It is found that high density permits ignition in a relatively small, moderately elongated plasma with a moderate magnetic field strength. Under these conditions, neutron wall loadings approximately 4 MW/m 2 must be tolerated. The sensitivity analysis with respect to impurity effects shows that impurity control will most likely be necessary to achieve the desired plasma conditions. The charge exchange sputtered impurities are found to have an important effect so that maintaining a low neutral density in the plasma is critical. If it is assumed that neutral beams will be used to heat the plasma to ignition, high energy injection is required (approximately 250 keV) when heating is accompished at full density. A scenario is outlined where the ignition temperature is established at low density and then the fueling rate is increased to attain ignition. This approach may permit beams with energies being developed for use in TFTR to be successfully used to heat a high density device of the type described here to ignition

  13. Tests of vacuum interrupters for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Warren, R.; Parsons, M.; Honig, E.; Lindsay, J.

    1979-04-01

    The Tokamak Fusion Test Reactor (TFTR) project at Princeton University requires the insertion of a resistor in an excited ohmic-heating coil circuit to produce a plasma initiation pulse (PIP). It is expected that the maximum duty for the switching system will be an interruption of 24 kA with an associated recovery voltage of 25 kV. Vacuum interrupters were selected as the most economical means to satisfy these requirements. However, it was felt that some testing of available systems should be performed to determine their reliability under these conditions. Two interrupter systems were tested for over 1000 interruptions each at 24 kA and 25 kV. One system employed special Westinghouse type WL-33552 interrupters in a circuit designed by LASL. This circuit used a commercially available actuator and a minimum size counterpulse bank and saturable reactor. The other used Toshiba type VGB2-D20 interrupters actuated by a Toshiba mechanism in a Toshiba circuit using a larger counterpulse bank and saturable reactor

  14. Proceedings of the workshop of three large tokamak cooperation on energy confinement scaling under intensive auxiliary heating, May 18 ∼ 20, 1992, Naka

    International Nuclear Information System (INIS)

    1992-09-01

    The workshop of three large tokamak cooperation W22 on 'Energy confinement scaling under intensive auxiliary heating' was held 18-20 May, 1992, at Naka Fusion Research Establishment. This proceedings compiles 14 synopses of contributions (5 from JET, 4 from JT-60, 3 from TFTR, and 1 each from DIII-D JFT-2M) and the summary of the workshop. Topic sections are ; (i) L-mode confinement and scaling, (ii) Confinement at high β P regimes, Supershots, High poloidal beta enhanced confinement mode etc., (iii) Confinement at various H-mode regimes and scaling (including the VH-mode), (iv) Characteristic time scales for present tokamak regimes, and (v) Theoretical comparison with experimental data. (author)

  15. Present status of Tokamak research

    International Nuclear Information System (INIS)

    Basu, Jayanta

    1991-01-01

    The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

  16. Simulation of lower hybrid current drive in enhanced reversed shear plasmas in the tokamak fusion test reactor using the lower hybrid simulation code

    International Nuclear Information System (INIS)

    Kaita, R.; Bernabei, S.; Budny, R.

    1996-01-01

    The Enhanced Reversed Shear (ERS) mode has already shown great potential for improving the performance of the Tokamak Fusion Test Reactor (TFTR) and other devices. Sustaining the ERS, however, remains an outstanding problem. Lower hybrid (LH) current drive is a possible method for modifying the current profile and controlling its time evolution. To predict its effectiveness in TFTR, the Lower Hybrid Simulation Code (LSC) model is used in the TRANSP code and the Tokamak Simulation Code (TSC). Among the results from the simulations are the following. (1) Single-pass absorption is expected in TFTR ERS plasmas. The simulations show that the LH current follows isotherms of the electron temperature. The ability to control the location of the minimum in the q profile (q min ) has been demonstrated by varying the phase velocity of the launched LH waves and observing the change in the damping location. (2) LH current drive can been used to sustain the q min location. The tendency of qmin to drift inward, as the inductive current diffuses during the formation phase of the reversed shear discharge, is prevented by the LH current driven at a fixed radial location. If this results in an expanded plasma volume with improved confinement as high power neutral beam injection is applied, the high bootstrap currents induced during this phase can then maintain the larger qmin radius. (3) There should be no LH wave damping on energetic beam particles. The values of perpendicular index of refraction in the calculations never exceed about 20, while ions at TFR injection energies are resonant with waves having values closer to 100. Other issues being addressed in the study include the LH current drive efficiency in the presence of high bootstrap currents, and the effect of fast electron diffusion on LH current localization

  17. Sheared Rotation Effects on Kinetic Stability in Enhanced Confinement Tokamak Plasmas, and Nonlinear Dynamics of Fluctuations and Flows in Axisymmetric Plasmas

    International Nuclear Information System (INIS)

    Beer, M.A.; Chance, M.S.; Hahm, T.S.; Lin, Z.; Rewoldt, G.; Tang, W.M.

    1997-01-01

    Sheared rotation dynamics are widely believed to have signficant influence on experimentally observed confinement transitions in advanced operating modes in major tokamak experiments, such as the Tokamak Fusion Test Reactor (TFTR) [D.J. Grove and D.M. Meade, Nuclear Fusion 25, 1167 (1985)], with reversed magnetic shear regions in the plasma interior. The high-n toroidal drift modes destabilized by the combined effects of ion temperature gradients and trapped particles in toroidal geometry can be strongly affected by radially sheared toroidal and poloidal plasma rotation. In previous work with the FULL linear microinstability code, a simplified rotation model including only toroidal rotation was employed, and results were obtained. Here, a more complete rotation model, that includes contributions from toroidal and poloidal rotation and the ion pressure gradient to the total radial electric field, is used for a proper self-consistent treatment of this key problem. Relevant advanced operating mode cases for TFTR are presented. In addition, the complementary problem of the dynamics of fluctuation-driven E x B flow is investigated by an integrated program of gyrokinetic simulation in annulus geometry and gyrofluid simulation in flux tube geometry

  18. New Monte Carlo results for the TFTR/Lithium Blanket Module system

    International Nuclear Information System (INIS)

    Engholm, B.A.

    1985-01-01

    Neutronics analysis results from Phase II of the TFTR Lithium Blanket Module (LBM) program are reported. Principal activities were analyses of new coverplate and protective plate designs; updating of the MCNP Monte Carlo model of TFTR/LBM; and performing new reference calculations for D-D and D-T plasmas. The new protective plate was found to reduce LBM responses by 20%. Updating the model included a new tally structure in which the LBM is divided into 92 volume elements corresponding to foil locations. A new version of the MCNP surface-source routine was used, along with the latest pointwise cross sections. All flux, tritium and foil responses are stored at NMFECC and are available for comparison with measurements, when the experimental program gets underway

  19. TFTR neutral beam D-T gas injection system operational experiences of the first two years

    International Nuclear Information System (INIS)

    Oldaker, M.E.; Lawson, J.E.; Stevenson, T.N.; Kamperschroer, J.H.

    1995-01-01

    The TFTR Neutral Beam Tritium Gas Injection System (TGIS) has successfully performed tritium operations since December 1993. TGIS operation has been reliable, with no leaks to the secondary containment to date. Notable operational problems include throughput leaks on fill, exit and piezoelectric valves. Repair of a TGIS requires replacement of the assembly, involving TFTR downtime and extensive purging, since the TGIS assembly is highly contaminated with residual tritium, and is located within secondary containment. Modifications to improve reliability and operating range include adjustable reverse bias voltage to the piezoelectric valves, timing and error calculation changes to tune the PLC and hardwired timing control, and exercising piezoelectric valves without actually pulsing gas prior to use after extended inactivity. A pressure sensor failure required the development of an open loop piezoelectric valve drive control scheme, using a simple voltage ramp to partially compensate for declining plenum pressure. Replacement of TGIS's have been performed, maintaining twelve system tritium capability as part of scheduled project maintenance activity

  20. Theory-based transport simulations of TFTR L-mode temperature profiles

    International Nuclear Information System (INIS)

    Bateman, G.

    1991-01-01

    The temperature profiles from a selection of TFTR L-mode discharges are simulated with the 1-1/2-D BALDUR transport code using a combination of theoretically derived transport models, called the Multi-Mode Model. The present version of the Multi-Mode Model consists of effective thermal diffusivities resulting from trapped electron modes and ion temperature gradient (η i ) modes, which dominate in the core of the plasma, together with resistive ballooning modes, which dominate in the periphery. Within the context of this transport model and the TFTR simulations reported here, the scaling of confinement with heating power comes from the temperature dependence of the η i and trapped electron modes, while the scaling with current comes mostly from resistive ballooning modes. 24 refs., 16 figs., 3 tabs

  1. Study of optically thin electron cyclotron emission from TFTR using a Michelson interferometer

    International Nuclear Information System (INIS)

    Stauffer, F.J.; Boyd, D.A.

    1986-01-01

    The TFTR Michelson interferometer, which is used as an electron temperature diagnostic, has a spectral range of 75-540 GHz. This range is adequate for measuring at least the first three cyclotron harmonics, and it spans both optically thick and thin portions of the ECE spectrum. During the most recent opening of the TFTR vacuum vessel, a concave, carbon reflector was installed on the back wall of the vessel, opposite the light collecting optic of the Michelson system. The reflector is designed to prevent the observation of optically thin ECE that originates from a location that is outside the field of view of the light collecting optic. If this is achieved, it should be possible to derive the electron density profile from measurements of either the extraordinary mode third harmonic or the ordinary mode second harmonic. An analysis of ECE spectra that have been measured before and after installation of the reflector is presented

  2. Analysis of erosion and transport of carbon impurity in the TFTR inner bumper limiter region

    International Nuclear Information System (INIS)

    Hua, T.Q.; Brooks, J.N.

    1992-01-01

    Carbon sputtering and transport on the TFTR inner graphite bumper limiter is investigated with the impurity transport code REDEP. Analysis is carried out for a series of ohmic discharges in TFTR. Predictions for Z eff in the core plasma agree well with in-situ experimental measurements. Run-away self-sputtering of carbon is predicted at low densities and high edge plasma temperatures when the limiter surface was purged of deuterium. Surface erosion and deposition is analyzed. In general, redeposition reduces the peak erosion by about a factor of five. Analysis is also carried out for a typical neutral beam heated discharge with a noncircular plasma. Spatial surface erosion and deposition profiles are compared qualitatively with beta backscattering measurements of metal deposition found on the limiter

  3. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

    1980-06-01

    The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

  4. Enhanced D-T supershot performance at high current using extensive lithium conditioning in TFTR

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Strachan, J.D.; Bell, M.G.; Scott, S.D.; Budny, R.; Bell, R.E.; Bitter, M.; Darrow, D.S.; Fredrickson, E.; Grek, B.

    1995-05-01

    A substantial improvement in supershot fusion plasma performance has been realized by combining the enhanced confinement due to tritium fueling with the enhanced confinement due to extensive Li conditioning of the TFTR limiter. This combination has resulted in not only significantly higher global energy confinement times than had previously been obtained in high current supershots, but also the highest ratio of central fusion output power to input power observed to date

  5. Analysis of momentum and impurity confinement in TFTR. [Annual report, 1989

    Energy Technology Data Exchange (ETDEWEB)

    1988-12-31

    The accomplishments to date of this research in collaboration with PPPL are the following: (1) full access capability to the TFTR data system has been achieved at Georgia Tech; (2) procedures to enable PPPL codes to be used in conjunction with ``in house`` programs for data analysis have been developed; (3) evaluation of the experimental data has been performed; and (4) a preliminary comparison of several momentum transport theories against experimental measurements has been performed.

  6. Structural analysis of equilibrium and ohmic heating coil assemblies for the TFTR

    International Nuclear Information System (INIS)

    Chattopadhyay, S.

    1975-10-01

    The structural adequacy of the equilibrium and ohmic heating coils and their support systems for the TFTR device has been investigated. The capability of the coils to span ribs of the support structure has been established. The support structure has been found to be effective in resisting the magnetic forces in the coils. The bands encircling the outboard coils and the band tensioning devices have been found to perform adequately. The analysis is based on October 1975 conceptual design

  7. Design and fabrication of an ion accelerator for TFTR-type neutral beam systems

    International Nuclear Information System (INIS)

    Paterson, J.A.; Duffy, T.J.; Haughian, J.M.; Biagi, L.A.; Yee, D.P.

    1977-10-01

    The design of the prototype 120-keV, 65-A, 0.5-sec ion accelerator for TFTR-type beam systems is described. Details of the manufacture of the constituent parts are given along with descriptions of the major components of the accelerator. Included are the molybdenum grid structures, molybdenum shields, stainless steel hats and the epoxy insulator. Specific manufacturing problems are discussed along with the results of tests to determine the voltage holding capabilities of the assembly

  8. Construction of the facility for the testing of the TFTR Neutral Beam Injector

    International Nuclear Information System (INIS)

    Haughian, J.; Lou, K.; Roth, D.

    1979-11-01

    The prototype for the TFTR Neutral Beam Injection System has been assembled at the Lawrence Berkeley Laboraory, and is presently under test. Some of the construction features of the shielding enclosure, the cryogenic supply system, control and computer area, and the auxiliary vacuum and utility supply system are described. In addition, the paper describes the target chamber, its beam dump and cryopanels, and the duct that connects the target chamber to the injector vessel

  9. MSC/NASTRAN ''expert'' techniques developed and applied to the TFTR poloidal field coils

    International Nuclear Information System (INIS)

    O'Toole, J.A.

    1986-01-01

    The TFTR poloidal field (PF) coils are being analyzed by PPPL and Grumman using MSC/NASTRAN as a part of an overall effort to establish the absolute limiting conditions of operation for TFTR. Each of the PF coils will be analyzed in depth, using a detailed set of finite element models. Several of the models developed are quite large because each copper turn, as well as its surrounding insulation, was modeled using solid elements. Several of the finite element models proved large enough to tax the capabilities of the National Magnetic Fusion Energy Computer Center (NMFECC), specifically disk storage space. To allow the use of substructuring techniques with their associated data bases for the larger models, it became necessary to employ certain infrequently used MSC/NASTRAN ''expert'' techniques. The techniques developed used multiple data bases and data base sets to divide each problem into a series of computer runs. For each run, only the data required was kept on active disk space, the remainder being placed in inactive ''FILEM'' storage, thus, minimizing active disk space required at any time and permitting problem solution using the NMFECC. A representative problem using the TFTR OH-1 coil global model provides an example of the techniques developed. The special considerations necessary to obtain proper results are discussed

  10. Tokamak reactor startup power

    International Nuclear Information System (INIS)

    Weldon, D.M.; Murray, J.G.

    1983-01-01

    Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

  11. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.

    1981-08-01

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  12. Theory of tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    White, R B [Princeton Univ., NJ (USA). Plasma Physics Lab.

    1989-01-01

    The book covers the consequences of ideal and resistive magnetohydrodynamics, these theories being responsible for most of what is well understood regarding the physics of tokamak discharges. The focus is on the description of equilibria, the linear and nonlinear theory of large scale modes, and single particle guiding center motion, including simple neoclassical effects. modern methods of general magnetic coordinates are used, and the student is introduced to the onset of chaos in Hamiltonian systems in the discussion of destruction of magnetic surfaces. Much of the book is devoted to the description of the limitations placed on tokamak operating parameters given by ideal and resistive modes, and current ideas about how to extend and optimize these parameters. (author). refs.; figs.

  13. Axisymmetric tokamak scapeoff transport

    International Nuclear Information System (INIS)

    Singer, C.E.; Langer, W.D.

    1982-08-01

    We present the first self-consistent estimate of the magnitude of each term in a fluid treatment of plasma transport for a plasma lying in regions of open field lines in an axisymmetric tokamak. The fluid consists of a pure hydrogen plasma with sources which arise from its interaction with neutral hydrogen atoms. The analysis and results are limited to the high collisionality regime, which is optimal for a gaseous neutralizer divertor, or to a cold plasma mantle in a tokamak reactor. In this regime, both classical and neoclassical transport processes are important, and loss of particles and energy by diamagnetic flow are also significant. The prospect of extending the analysis to the lower collisionality regimes encountered in many existing experiments is discussed

  14. Density limits in Tokamaks

    International Nuclear Information System (INIS)

    Tendler, M.

    1984-06-01

    The energy loss from a tokamak plasma due to neutral hydrogen radiation and recycling is of great importance for the energy balance at the periphery. It is shown that the requirement for thermal equilibrium implies a constraint on the maximum attainable edge density. The relation to other density limits is discussed. The average plasma density is shown to be a strong function of the refuelling deposition profile. (author)

  15. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.

    1984-05-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

  16. Modular tokamak magnetic system

    International Nuclear Information System (INIS)

    Yang, T.F.

    1988-01-01

    This patent describes a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of the vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein the magnetic coil is wound around the toroidal field coil such that the ohmic heating coil enclosed the toroidal field coil

  17. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.; California Univ., Los Angeles

    1984-01-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scrape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this 'Z-mode' of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described. (orig.)

  18. TPX tokamak construction management

    International Nuclear Information System (INIS)

    Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.

    1995-01-01

    A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly

  19. Proposed experiment to investigate use of heated optical fibers for tokamak diagnostics during D-T discharges

    International Nuclear Information System (INIS)

    Tighe, W.; Morgan, P.; Griscom, D.; Adler, H.; Cylinder, D.; Johnson, D.; Palladino, D.; Ramsey, A.

    1995-02-01

    A collaborative JET/TFTR study has been undertaken to investigate attenuation and luminescence effects due to neutron irradiation of optical fibers heated to 400 degrees C. It is expected that a significant improvement in fiber behavior will be observed due to thermal annealing. This technique may be important for use in fiber-related, tokamak diagnostics exposed to high neutron flux. The study will make use of aluminum jacketed, 600 μm diameter, all silica (F-doped cladding) fibers in lengths of 150 m. The fibers are prepared in 1 foot coils. Of the coils to be irradiated, one is heated constantly to 400 degrees C, a second is not heated, and a third is heated periodically. A fourth fiber coil is not to be irradiated. Spectrally and temporally resolved transmission and luminescence data under neutron irradiation during D-T discharges on TFTR will be obtained. An investigation of permanent and short term effects will be made. Experimental details along with initial results will be presented

  20. Design of Fire/Gas Penetration Seals and fire exposure tests for Tokamak Fusion Test Reactor experimental areas

    International Nuclear Information System (INIS)

    Cavalluzzo, S.

    1983-01-01

    A Fire/Gas Penetration Seal is required in every penetration through the walls and ceilings into the Test Cell housing the Tokamak Fusion Test Reactor (TFTR), as well as other adjacent areas to protect the TFTR from fire damage. The penetrations are used for field coil lead stems, diagnostics systems, utilities, cables, trays, mechanical devices, electrical conduits, vacuum liner, air conditioning ducts, water pipes, and gas pipes. The function of the Fire/Gas Penetration Seals is to prevent the passage of fire and products of combustion through penetrations for a period of time up to three hours and remain structurally intact during fire exposure. The Penetration Seal must withstand, without rupture, a fire hose water stream directed at the hot surface. There are over 3000 penetrations ranging in size from several square inches to 100 square feet, and classified into 90 different types. The material used to construct the Fire/Gas Penetration Seals consist of a single and a two-component room temperature vulcanizing (RTV) silicone rubber compound. Miscellaneous materials such as alumina silica refractory fibers in board, blanket and fiber forms are also used in the construction and assembly of the Seals. This paper describes some of the penetration seals and the test procedures used to perform the three-hour fire exposure tests to demonstrate the adequacy of the seals

  1. Status of the tokamak program

    Science.gov (United States)

    Sheffield, J.

    1981-08-01

    For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

  2. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1994-01-01

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  3. The density limit in Tokamaks

    International Nuclear Information System (INIS)

    Alladio, F.

    1985-01-01

    A short summary of the present status of experimental observations, theoretical ideas and understanding of the density limit in tokamaks is presented. It is the result of the discussion that was held on this topic at the 4th European Tokamak Workshop in Copenhagen (December 4th to 6th, 1985). 610 refs

  4. Radiation analysis of the CIT (Compact Ignition Tokamak) pellet injector system and its impact on personnel access

    Energy Technology Data Exchange (ETDEWEB)

    Selcow, E.C.; Stevens, P.N.; Gomes, I.C.; Gomes, L.M.

    1987-01-01

    Conceptual design of the Compact Ignition Tokamak (CIT) is near completion. This short-pulse ignition experiment is planned to follow the operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The high neutron wall loadings, /approximately/4-5 MW/m/sup 2/, associated with the operation of this device require that neutronics-related issues be considered in the overall system design. Radiation shielding is required for the protection of device components and personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure, and the entire experiment is housed in a circular test cell facility with a radius of /approximately/12 m. The most critical radiation concern in the CIT design process relates to the numerous penetrations in the device. This paper discusses the impact of a major penetration on the design and operations of the CIT pellet injection system. The pellet injector is a major component, which has a line-of-sight penetration through the igloo and test cell wall. All current options for maintenance of the injector require personnel access. A nuclear analysis has been performed to determine the feasibility of hands-on access. Results indicate that personnel access to the pellet injector glovebox is possible. 10 refs., 3 figs., 3 tabs.

  5. Thermal and nonthermal electron cyclotron emission by high-temperature tokamak plasmas

    International Nuclear Information System (INIS)

    Airoldi, A.; Ramponi, G.

    1997-01-01

    An analysis of the electron cyclotron emission (ECE) spectra emitted by a high-temperature tokamak plasma in the frequency range of the second and third harmonic of the electron cyclotron frequency is made, both in purely Maxwellian and in non-Maxwellian cases (i.e., in the presence of a current-carrying superthermal tail). The work is motivated mainly by the experimental observations made in the supershot plasmas of the Tokamak Fusion Test Reactor (TFTR), where a systematic disagreement is found between the T e measurements by second-harmonic ECE and Thomson scattering. We show that, by properly taking into account the overlap of superthermals-emitted third harmonic with second-harmonic bulk emission, the radiation temperature observed about the central frequency of the second harmonic may be enhanced up to 30%endash 40% compared to the corresponding thermal value. Moreover we show that, for parameters relevant to the International Thermonuclear Experimental Reactor (ITER) with T e (0)>7 keV, the overlap between the second and the downshifted third harmonic seriously affects the central plasma region, so that the X-mode emission at the second harmonic becomes unsuitable for local T e measurements. copyright 1997 American Institute of Physics

  6. Charge-exchange and fusion reaction measurements during compression experiments with neutral beam heating in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Kaita, R.; Heidbrink, W.W.; Hammett, G.W.

    1986-04-01

    Adiabatic toroidal compression experiments were performed in conjunction with high power neutral beam injection in the Tokamak Fusion Test Reactor (TFTR). Acceleration of beam ions to energies nearly twice the injection energy was measured with a charge-exchange neutral particle analyzer. Measurements were also made of 2.5 MeV neutrons and 15 MeV protons produced in fusion reactions between the deuterium beam ions and the thermal deuterium and 3 He ions, respectively. When the plasma was compressed, the d(d,n) 3 He fusion reaction rate increased a factor of five, and the 3 He(d,p) 4 He rate by a factor of twenty. These data were simulated with a bounce-averaged Fokker-Planck program, which assumed conservation of angular momentum and magnetic moment during compression. The results indicate that the beam ion acceleration was consistent with adiabatic scaling

  7. Alpha-driven magnetohydrodynamics (MHD) and MHD-induced alpha loss in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Chang, Z.; Nazikian, R.; Fu, G.Y.

    1997-02-01

    Alpha-driven toroidal Alfven eigenmodes (TAEs) are observed as predicted by theory in the post neutral beam phase in high central q (safety factor) deuterium-tritium (D-T) plasmas in the Tokamak Fusion Test Reactor (TFTR). The mode location, poloidal structure and the importance of q profile for TAE instability are discussed. So far no alpha particle loss due to these modes was detected due to the small mode amplitude. However, alpha loss induced by kinetic ballooning modes (KBMs) was observed in high confinement D-T discharges. Particle orbit simulation demonstrates that the wave-particle resonant interaction can explain the observed correlation between the increase in alpha loss and appearance of multiple high-n (n ≥ 6, n is the toroidal mode number) modes

  8. Estimated neutron-activation data for TFTR. Part II. Biological dose rate from sample-materials activation

    International Nuclear Information System (INIS)

    Ku, L.; Kolibal, J.G.

    1982-06-01

    The neutron induced material activation dose rate data are summarized for the TFTR operation. This report marks the completion of the second phase of the systematic study of the activation problem on the TFTR. The estimations of the neutron induced activation dose rates were made for spherical and slab objects, based on a point kernel method, for a wide range of materials. The dose rates as a function of cooling time for standard samples are presented for a number of typical neutron spectrum expected during TFTR DD and DT operations. The factors which account for the variations of the pulsing history, the characteristic size of the object and the distance of observation relative to the standard samples are also presented

  9. Tokamak instrumentation and controls

    International Nuclear Information System (INIS)

    Becraft, W.R.; Bettis, E.S.; Houlberg, W.A.; Onega, R.J.; Stone, R.S.

    1979-02-01

    The three areas of study emphasis to date are: (1) Physics implications for controls, (2) Computer simulation, and (3) Shutdown/aborts. This document reports on the FY 78 efforts (the first year of these studies) to address these problems. Transient scenario options for the startup of a tokamak are developed, and the implications for the control system are discussed. This document also presents a hybrid computer simulation (analog and digital) of the Impurity Study Experiment (ISX-B) which is now being used for corroborative controls investigations. The simulation will be expanded to represent a TNS/ETF machine

  10. Demonstration tokamak power plant

    International Nuclear Information System (INIS)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System

  11. Maximum entropy tokamak configurations

    International Nuclear Information System (INIS)

    Minardi, E.

    1989-01-01

    The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)

  12. Tokamak Fusion Test Reactor. Final conceptual design report

    International Nuclear Information System (INIS)

    1976-02-01

    The TFTR is the first U.S. magnetic confinement device planned to demonstrate the fusion of D-T at reactor power levels. This report addresses the physics objectives and the engineering goals of the TFTR project. Technical, cost, and schedule aspects of the project are included

  13. Parametric analysis of neutron streaming through major penetrations in the 0.914 m TFTR test cell floor

    International Nuclear Information System (INIS)

    Ku, L.P.; Liew, S.L.; Kolibal, J.G.

    1985-09-01

    Neutron streaming through penetrations in the 0.914 m TFTR test cell floor has two distinct features: (1) the oblique angle of incidence; and (2) the high order of anisotropy in the angular distribution for incident neutrons with energies > 10 keV. The effects of these features on the neutron streaming into the TFTR basement were studied parametrically for isolated penetrations. Variations with respect to the source energies, angular distributions, and sizes of the penetrations were made. The results form a data base from which the spatial distribution of the neutron flux in the basement due to multiple penetrations may be evaluated

  14. Topology of tokamak orbits

    International Nuclear Information System (INIS)

    Rome, J.A.; Peng, Y.K.M.

    1978-09-01

    Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having β's of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher β (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower β case. The differences indicate the confinement of additional high energy (v → c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well

  15. ITER tokamak device

    International Nuclear Information System (INIS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-01-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  16. Dust Measurements in Tokamaks

    International Nuclear Information System (INIS)

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-01-01

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 (micro)m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics

  17. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.

    1991-02-01

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least κ sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  18. Analysis of electron cyclotron emission spectra of high electron temperature, supershot plasmas in TFTR

    International Nuclear Information System (INIS)

    Taylor, G.; Arunasalam, V.; Efthimion, P.C.; Grek, B.

    1993-01-01

    A primary objective of the TFTR program since 1986 has been the study and optimization of deuterium Supershot plasmas. These plasmas are predominantly heated by 90-110 keV neutral deuterium beams (P NBI /P OH >30), central ion temperatures are ∝30 keV and central electron temperatures from ECE (T ECE ) often exceed 10 keV. Central electron temperature data measured with a TV Thomson scattering (TVTS) system (T TVTS ) during the period 1987-1990 have been compared with data from three different ECE instruments on TFTR. Although T ECE ∝T TVTS for temperatures below 6 keV, there is a systematically increasing disagreement at higher electron temperatures, with T ECE ∝1.2 T TVTS for T TVTS in the range 9-10 keV. Recent theoretical work on the ECE radiation temperature of non-equilibrium plasmas indicates that for a bi-Maxwellian electron velocity distribution with a ratio of tail to bulk electron density η, a bulk temperature T b , and a hot tail temperature T h , the perpendicular ECE radiation temperature is given by T ECE ∝T b {1+η(T h /T b )}, for η ECE would be enhanced over T TVTS by a factor which depends on η and T h . This paper investigates whether the discrepancy between T TVTS and T ECE seen in TFTR Supershots at high electron temperatures is due to the presence of a hot electron tail component. The extraordinary mode ECE spectrum at the second, third and fourth harmonics is measured on the horizontal midplane by an absolutely calibrated ECE Michelson interferometer. This ECE spectrum is compared with the output from a time-independent transport code with relativistic opacity which solves the three-dimensional ECE radiation transport in a toroidally symmetric, two-dimensional geometry and uses measured electron density and temperature profiles from the TVTS system. (orig.)

  19. Modelling dust transport in tokamaks

    International Nuclear Information System (INIS)

    Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.

    2008-01-01

    The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)

  20. Fusion power production from TFTR plasmas fueled with deuterium and tritium

    International Nuclear Information System (INIS)

    Strachan, J.D.; Adler, H.; Alling, P.

    1994-03-01

    Peak fusion power production of 6.2 ± 0.4 MW has been achieved in TFTR plasmas heated by deuterium and tritium neutral beams at a total power of 29.5 MW. These plasmas have an inferred central fusion alpha particle density of 1.2 x 10 17 m -3 without the appearance of either disruptive MHD events or detectable changes in Alfven wave activity. The measured loss rate of energetic alpha particles agreed with the approximately 5% losses expected from alpha particles which are born on unconfined orbits