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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.)

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

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)

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

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

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

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

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.

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

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

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

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

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

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)

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

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

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

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

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.)

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

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

Scheduling and control of programs in the CICADA computer system for TFTR

International Nuclear Information System (INIS)

Christianson, G.B.; Sherrick, M.; Stark, W.

1983-01-01

The CICADA computer system is a network of super minicomputers for data acquisition, reduction, display and device control on the TFTR experiment. Human interaction with the CICADA system is accomplished by consoles offering color displays with a high degree of operator interaction with the computer system, and by interactive terminals offering a limited degree of operator interaction. The authors describe the software packages and data bases used to control the consoles and provide communication to an application program attached to a console. The console control program in its turn schedules the execution of control programs that furnish operating system interfaces to applications programs executing at a console or terminal (in response to human activation), or upon activation by scheduling tasks. The control programs handle functions such as applications task activation. They discuss the implementation of the control programs in the multi-computer CICADA environment, and describe the control program data bases. In a TFTR shot cycle, applications programs must be scheduled for activation at operator-defined stages, in order to accomplish the arming of hardware devices, acquisition of summary waveform data, and the acquisition and archiving of raw data. The authors describe the software tasks and data bases of the CICADA event system, which accomplishes this scheduling of applications task activation in conjunction with the control programs. Future enhancements planned for the console control, program control, and event systems are outlined

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

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

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

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

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)

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

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

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

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

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

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

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

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

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

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%

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

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

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

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

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)

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.

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

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

Non-superconducting magnet structures for near-term, large fusion experimental devices

International Nuclear Information System (INIS)

File, J.; Knutson, D.S.; Marino, R.E.; Rappe, G.H.

1980-10-01

This paper describes the magnet and structural design in the following American tokamak devices: the Princeton Large Torus (PLT), the Princeton Divertor Experiment (PDX), and the Tokamak Fusion Test Reactor (TFTR). The Joint European Torus (JET), also presented herein, has a magnet structure evolved from several European programs and, like TFTR, represents state of the art magnet and structure design

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.

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

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

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.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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)

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.)

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Ion cyclotron emission due to collective instability of fusion products and beam ions in TFTR and JET

International Nuclear Information System (INIS)

Dendy, R.O.; Clements, K.G.; Lashmore-Davies, C.N.; Cottrell, G.A.; Majeski, R.; Cauffman, S.

1995-06-01

Ion cyclotron emission (ICE) has been observed from neutral beam-heated TFTR and JET tritium experiments at sequential cyclotron harmonics of both fusion products and beam ions. The emission originates from the outer mid-plane plasma, where fusion products and beam ions are likely to have a drifting ring-type velocity-space distribution which is anisotropic and sharply peaked. Fusion product-driven ICE in both TFTR and JET can be attributed to the magnetoacoustic cyclotron instability, which involves the excitation of obliquely propagating waves on the fast Alfven/ion Bernstein branch at cyclotron harmonics of the fusion products. Differences between ICE observations in JET and TFTR appear to reflect the sensitivity of the instability growth rate to the ratio υ birth /c A , where υ birth is the fusion product birth speed and c A is the local Alfven speed:for fusion products in the outer midplane edge of TFTR, υ birth A ; for alpha-particles in the outer midplane edge of JET, the opposite inequality applies. If sub-Alfvenic fusion products are isotropic or have undergone even a moderate degree of thermalization, the magnetoacoustic instability cannot occur. In contrast, the super-Alfvenic alpha-particles which are present in the outer mid-plane of JET can drive the magnetoacoustic cyclotron instability even if they are isotropic or have a relatively broad distribution of speeds. These conclusions may account for the observation that fusion product-driven ICE in JET persists for longer than fusion product-driven ICE in TFTR. (Author)

Selected topics on surface effects in fusion devices: neutral-beam injectors and beam-direct converters

International Nuclear Information System (INIS)

Kaminsky, M.

1978-01-01

Neutral-beam injectors are being used for the heating and fueling of plasmas in existing devices such as PLT (Princeton), ISX (Oak Ridge) and 2XIIB (Lawrence Livermore Laboratory) and will be used in devices such as TFTR (Princeton), MX (Livermore) and Doublet III (Gulf Atomic). For example, TFTR has been designed to receive a total of 20 MW of 120-keV deuterium atoms in pulses of 0.5-sec duration from 12 neutral beam injectors; for the MX experiment it is planned to inject a total of 750A (equivalent) of deuterium atoms with a mean energy of 56 keV in 0.5-sec pulses. The interaction of energetic deuterium atoms with exposed surfaces of device components such as beam dumps, beam-direct-convertors collectors, beam calorimeters, and armor plates, cause a variety of surface effects which affect deleteriously the operation of such devices. Some of the major effects will be discussed

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

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

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

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

[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

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

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.

A comparison of hydrogen vs. helium glow discharge effects on fusion device first-wall conditioning

International Nuclear Information System (INIS)

Dylla, H.F.

1989-09-01

Hydrogen- and deuterium-fueled glow discharges are used for the initial conditioning of magnetic fusion device vacuum vessels following evacuation from atmospheric pressure. Hydrogenic glow discharge conditioning (GDC) significantly reduces the near-surface concentration of simple adsorbates, such as H 2 O, CO, and CH 4 , and lowers ion-induced desorption coefficients by typically three orders of magnitude. The time evolution of the residual gas production observed during hydrogen-glow discharge conditioning of the carbon first-wall structure of the TFTR device is similar to the time evolution observed during hydrogen GDC of the initial first-wall configuration in TFTR, which was primarily stainless steel. Recently, helium GDC has been investigated for several wall-conditioning tasks on a number of tokamaks including TFTR. Helium GDC shows negligible impurity removal with stainless steel walls. For impurity conditioning with carbon walls, helium GDC shows significant desorption of H 2 O, CO, and CO 2 ; however, the total desorption yield is limited to the monolayer range. In addition, helium GDC can be used to displace hydrogen isotopes from the near-surface region of carbon first-walls in order to lower hydrogenic retention and recycling. 38 refs., 6 figs

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

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

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

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.

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

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

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

Ion cyclotron emission due to collective instability of fusion products and beam ions in TFTR and JET

International Nuclear Information System (INIS)

Dendy, R.O.; McClements, K.G.; Lashmore Davies, C.N.; Cottrell, G.A.; Majeski, R.; Cauffman, S.

1995-01-01

Ion cyclotron emission (ICE) has been observed from neutral beam heated TFTR and JET tritium experiments at sequential cyclotron harmonics of both fusion products and beam ions. The emission originates from the outer midplane plasma, where fusion products and beam ions are likely to have a drifting ring-type velocity-space distribution that is anisotropic and sharply peaked. Fusion product driven ICE can be attributed to the magnetoacoustic cyclotron instability, which involves the excitation of obliquely propagating waves on the fast Alfven/ion Bernstein branch at cyclotron harmonics of the fusion products. Differences between ICE observations in JET and TFTR appear to reflect the sensitivity of the instability growth rate to the ratio υ birth /c A , where υ birth is the fusion product birth speed and c A is the local Alfven speed: for fusion products in the outer midplane edge of TFTR supershots, υ birth A ; for alpha particles in the outer midplane edge of JET, the opposite inequality applies. If sub-Alfvenic fusion products are isotropic or have undergone even a moderate degree of thermalization, the magnetoacoustic instability cannot occur. In contrast, the super-Alfvenic alpha particles that are present in the outer midplane of JET can drive the magnetoacoustic cyclotron instability even if they are isotropic or have a relatively broad distribution of speeds. These conclusions may account for the observation that fusion product driven ICE in JET persists for longer than fusion product driven ICE in TFTR. A separate mechanism is proposed for the excitation of beam driven ICE in TFTR: electrostatic ion cyclotron harmonic waves, supported by strongly sub-Alfvenic beam ions, can be destabilized by a low concentration of such ions with a very anrrow spread of velocities in the parallel direction. 25 refs, 14 figs

Transport simulations of ohmic TFTR experiments with profile-consistent microinstability-based models for chi/sub e/ and chi/sub i/

International Nuclear Information System (INIS)

Redi, M.H.; Tang, W.M.; Efthimion, P.C.; Mikkelsen, D.R.; Schmidt, G.L.

1987-03-01

Transport simulations of ohmically heated TFTR experiments with recently developed profile-consistent microinstability models for the anomalous thermal diffusivities, chi/sub e/ and chi/sub i/, give good agreement with experimental data. The steady-state temperature profiles and the total energy confinement times, tau/sub e/, were found to agree for each of the ohmic TFTR experiments simulated, including three high radiation cases and two plasmas fueled by pellet injection. Both collisional and collisionless models are tested. The trapped-electron drift wave microinstability model results are consistent with the thermal confinement of large plasma ohmic experiments on TFTR. We also find that transport due to the toroidal ion temperature gradient (eta/sub i/) modes can cause saturation in tau/sub E/ at the highest densities comparable to that observed on TFTR and equivalent to a neoclassical anomaly factor of 3. Predictions based on stabilized eta/sub i/-mode-driven ion transport are found to be in agreement with the enhanced global energy confinement times for pellet-fueled plasmas. 33 refs., 26 figs., 4 tabs

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

Measurement of internal magnetic field pitch using Li pellet injection on TFTR (invited)

International Nuclear Information System (INIS)

Terry, J.L.; Marmar, E.S.; Howell, R.B.; Bell, M.; Cavallo, A.; Fredrickson, E.; Ramsey, A.; Schmidt, G.L.; Stratton, B.; Taylor, G.; Mauel, M.E.

1990-01-01

A diagnostic technique which measures the direction of the internal magnetic field pitch angle has been used successfully on TFTR. The technique requires the injection of high-speed Li pellets. The magnetic field direction is measured by observing the polarization direction of the intense visible line emission from Li + (λ∼5485 A, 1s2p 3 P 0,1,2 →1s2s 3 S 0 ) in the pellet ablation cloud. The presence of the large (primarily toroidal) magnetic field causes the line to be split due to the Zeeman effect, and the unshifted π component is polarized with its polarization direction parallel to the local magnetic field. In devices with sufficiently strong fields (B approx-gt 4.5 T), the Zeeman splitting of the line is large enough, relative to the linewidth of each Zeeman component, that enough residual polarization remains. Because the pellet moves about 1 cm before the Li + is ionized (τ ionization approx-lt 10 μs), the time history of the polarization direction (as the pellet penetrates from the outside toward the plasma center) yields the local magnetic field direction. In the TFTR experiment, spatial resolution of the measurement is typically ∼7 cm, limited by the requirement that a large number of photons must be collected in order to make the measurement of the polarization angle. Typically, the pitch of the field is measured with an accuracy of ±0.01 rad, limited by the photon statistics. The measurements of the internal field pitch angle, combined with external magnetic measurements, have been used in a code which finds the solution of the Grad--Shafranov equation, yielding the equilibrium which is the best fit to the measured inputs

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

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

Interrupter and hybrid-switch testing for fusion devices

International Nuclear Information System (INIS)

Parsons, W.M.; Warren, R.W.; Honig, E.M.; Lindsay, J.D.G.; Bellamo, P.; Cassel, R.L.

1979-01-01

This paper discusses recent and ongoing switch testing for fusion devices. The first part describes testing for the TFTR ohmic-heating circuit. In this set of tests, which simulated the stresses produced during a plasma initiation pulse, circuit breakers were required to interrupt a current of 24 kA with an associated recovery voltage of 25 kV. Two interrupter systems were tested for over 1000 operations each, and both appear to satisfy TFTR requirements. The second part discusses hybrid-switch development for superconducting coil protection. These switching systems must be capable of carrying large currents on a continuous basis as well as performing interruption duties. The third part presents preliminary results on an early-counterpulse technique applied to vacuum interrupters. Implementation of this technique has resulted in large increases in interruptible current as well as a marked reduction in contact erosion

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

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

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

Transport simulations of TFTR experiments to test theoretical models for χe and χi

International Nuclear Information System (INIS)

Redi, M.H.; Bateman, G.

1990-08-01

1-1/2-d BALDUR transport code predictions using recent theoretically-based models for thermal and particle transport are compared to measured profiles of electron plasma density and electron and ion temperatures for TFTR ohmic, L-mode and supershot discharges. The profile consistent drift wave model is found to overestimate ion temperatures at high heating powers, so that a third mode or loss process is needed in addition to drift wave transport (TEM, η i ) and an edge loss model. None of several versions of local multiple mode models, using the 1989 Carreras-Diamond resistive ballooning model, gives T e , T i within 20% for all three TFTR regimes studied. 36 refs., 7 figs., 4 tabs

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

Layout of the manipulator-arm (boom) for the TFTR fusion reactor (Princeton, USA) under UHV-conditions

International Nuclear Information System (INIS)

Klaubert, J.

1987-01-01

This presentation shows the main criteria for the layout of the manipulator - arm and the antechamber - vessel of the TFTR - FUSION - REACTOR at Princeton University, PLASMA PHYSICS LABORATORY (USA). The main problem during layout of a manipulator system like the TFTR - Boom has been the limitation of the vertical deflections due to deadweight of the construction. The design problem is rather a deformation problem and a problem of stability than a stress problem. The way of optimizing the ratio between stiffness and deadweight is the most important part during the complete design - process. Additional earthquake requirements need further investigations for a satisfying layout (horizontal forces, weak-axis of moment of inertia). The details of the construction (welding, connections etc.) have to be designed in respect to UHV - requirements --> no holes and no fillet welds (outgasing - rate.) are allowed. All weldings have to be designed as bevel-welds. This manipulator system is designed for working in a plane system (two degrees of freedom). A manipulator system with the same operating capabilities in a three degree of freedom system needs larger cross sections for the different beam-elements than those of the discussed TFTR - BOOM

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

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

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

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

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

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

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

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)

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)

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

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

Achieving high fusion reactivity in high poloidal beta discharges in TFTR

International Nuclear Information System (INIS)

Manuel, M.E.; Navratil, G.A.; Sabbagh, S.A.; Batha, S.; Bell, M.G.; Bell, R.; Budny, R.V.; Bush, C.E.; Cavallo, A.; Chance, M.S.; Cheng, C.Z.; Efthimion, P.C.; Fredrickson, E.D.; Fu, G.Y.; Hawryluk, R.J.; Janos, A.C.; Jassby, D.L.; Levinton, F.; Mikkelsen, D.R.; Manickam, J.; McCune, D.C.; McGuire, K.M.; Medley, S.S.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Park, H.K.; Ramsey, A.T.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Wieland, R.M.; Yamada, M.; Zarnstorff, M.C.: Zweben, S.; Kesner, J.; Marmar, E.; Snipes, J.; Terry, J.

1993-04-01

High poloidal beta discharges have been produced in TFTR that achieved high fusion reactivities at low plasma currents. By rapidly decreasing the plasma current just prior to high-power neutral beam injection, relatively peaked current profiles were created having high l i > 2, high Troyon-normalized beta, βN > 3, and high poloidal beta. β p ≥ 0.7 R/a. The global energy confinement time after the current ramp was comparable to supershots, and the combination of improved MHD stability and good confinement produced a new high εβ p high Q DD operating mode for TFTR. Without steady-state current profile control, as the pulse lengths of high βp discharges were extended, l i decreased, and the improved stability produced immediately after by the current ramp deteriorated. In four second, high εβ p discharges, the current profile broadened under the influence of bootstrap and beam-drive currents. When the calculated voltage throughout the plasma nearly vanished, MHD instabilities were observed with β N as low as 1.4. Ideal MHD stability calculations showed this lower beta limit to be consistent with theoretical expectations

Measurement of loss of DT fusion products using scintillator detectors in TFTR

International Nuclear Information System (INIS)

Darrow, D.S.; Herrmann, H.W.; Johnson, D.W.; Marsala, R.J.; Palladino, R.W.; Zweben, S.J.

1995-03-01

A poloidal array of MeV ion loss probes previously used to measure DD fusion product loss has been upgraded to measure the loss of alpha particles from DT plasmas in TFTR. The following improvements to the system have been made in preparation for the use of tritium in TFTR: (1) relocation of detectors to a neutronshielded enclosure in the basement to reduce neutron-induced background signals; (2) replacement of ZnS:Cu (P31) scintillators in the probes with the Y 3 Al 5 0 12 :Ce(P46) variety to minimize damage and assure linearity at the fluxes anticipated from DT plasmas; and (3) shielding of the fiber optic bundles which carry the fight from the probes to the detectors to reduce neutron- and gamma-induced light within them. In addition to the above preparations, the probes have been absolutely calibrated for alpha particles by using the Van de Graaf accelerator at Los Alamos National Laboratory. Alpha particle losses from DT plasmas have been observed, and losses at the detector 901 below the midplane are consistent with first orbit loss

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

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

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

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

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

Eddy current analysis in fusion devices

International Nuclear Information System (INIS)

Turner, L.R.

1988-06-01

In magnetic fusion devices, particularly tokamaks and reversed field pinch (RFP) experiments, time-varying magnetic fields are in intimate contact with electrically conducting components of the device. Induced currents, fields, forces, and torques result. This note reviews the analysis of eddy current effects in the following systems: Interaction of a tokamak plasma with the eddy currents in the first wall, blanket, and shield (FWBS) systems; Eddy currents in a complex but two-dimensional vacuum vessel, as in TFTR, JET, and JT-60; Eddy currents in the FWBS system of a tokamak reactor, such as NET, FER, or ITER; and Eddy currents in a RFP shell. The cited studies are chosen to be illustrative, rather than exhaustive. 42 refs

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

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

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

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

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)

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).

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

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.

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

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

Loss of alpha-like MeV fusion products from TFTR

International Nuclear Information System (INIS)

Zweben, S.J.; Boivin, R.L.; Diesso, M.; Hayes, S.E.; Hendel, H.W.; Park, H.; Strachan, J.D.

1990-03-01

A detailed comparison between the observed and expected loss of alpha-like MeV fusion products in TFTR is presented. The D-D fusion products (mainly the 1 MeV triton) were measured with an 2-D imaging scintillation detector. The expected first-orbit loss was calculated with a simple Lorentz orbit code. In almost all cases the measured loss was consistent with the expected first-orbit loss model. Exceptions are noted for small major radius plasmas and during strong MHD activity. 37 refs., 28 figs

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

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.))

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

Measurement of the nonaxisymmetric heat load distribution on the first wall of TFTR due to locked modes

International Nuclear Information System (INIS)

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

1992-01-01

The first wall of TFTR is covered in large part (23%) by an inner-wall bumper limiter which is the primary power handling structure in TFTR. The limiter is comprised of more than 2000 tiles, and is instrumented with a large number (>100) of thermocouples in a two-dimensional (2D) array, primarily for protection of the wall. While only about 5% of the tiles are monitored, this thermocouple system is nevertheless capable of mapping details in the nonaxisymmetric, as well as symmetric, heat load patterns encountered under different conditions. In particular, helical heating patterns are observed in discharges which have locked modes. The helical patterns clearly match the expected trajectories based on the m/n mode numbers obtained from Mirnov coils (m/n=2/1 and 4/1), so that the thermocouple system can and was used to identify the existence and mode number of a locked mode. While TFTR discharges rarely suffer from locked modes, locked modes always alter the heating pattern. The locked modes are found to very significantly redistribute the heat load for both ohmic and NBI heated discharges. Locked modes can make what were the coldest areas into the hottest areas, and vice versa. Locked modes also can alter the heat pattern resulting from the frequent disruptions which occur as a result of a locked mode

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

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

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

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

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.)

Simulation of mode converted ion Bernstein wave - beam deuteron interactions on TFTR

Science.gov (United States)

Herrmann, Mark; Fisch, Nathaniel

1998-11-01

Experiments on TFTR have documented strong interactions between mode converted ion Bernstein waves (MCIBW) and beam deuterons(D. S. Darrow et al.), Nucl. Fusion 36, 509 (1996).^,(N. J. Fisch et al.), IAEA, Vol. 1, p. 271 (1996). This is of particular interest in the study of α channelling, since the most promising scenarios(M. C. Herrmann and N. J. Fisch, Phys. Rev. Lett. 79), 1495 (1997). rely on a suitable combination of MCIBW and Alfvén eigenmodes to achieve the cooling of the α particles. Collisional effects, realistic wave fields, and a detailed model of the wave-particle interaction have been added to the Monte Carlo simulations which are used to simulate α channelling in order to model TFTR experiments(M. C. Herrmann, Ph.D. thesis, Princeton University, 1998.). The results are found to be in qualitative agreement with the data. In addition, the simulation is used, in conjunction with the data, to demonstrate the existence of the k_\\|-flip of the MCIBW, and to infer a diffusion coefficient for the beam deuterons interacting with the wave. This diffusion coefficient significantly exceeds what would be expected on the basis of quasilinear theory with the fields specified by 1 D ray tracing of the MCIBW.

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

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

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.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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%

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

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

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

The effect of ICRF antenna phasing on metal impurities in TFTR

International Nuclear Information System (INIS)

Stevens, J.E.; Bush, C.; Colestock, P.L.; Oak Ridge National Lab., TN; AN Ukrainskoj SSR, Kharkov

1989-07-01

ICRF power levels of up to 2.8 MW were achieved during the 1988 experimental run on TFTR. Metal impurity concentrations (Ti, Cr, Fe, Ni) and Z eff were monitored during ICRF heating by x-ray pulse height analysis and uv spectroscopy. Antenna phasing was the key variable affecting ICRF performance. No increase in metallic impurities was observed for P rf approx lt 2.8 MW with the antenna straps 0-Π, while a measurable increase in titanium (Faraday screen material) was observed for P rf approx gt 1.0 MW with 0-0 phasing. 18 refs., 8 figs

MeV ion loss during 3He minority heating in TFTR

International Nuclear Information System (INIS)

Zweben, S.J.; Hammett, G.; Boivin, R.; Phillips, C.; Wilson, R.

1992-01-01

The loss of MeV ions during 3 He ICRH minority heating experiments has been measured using scintillator detectors near the wall of TFTR. The observed MeV ion losses to the bottom (90 degrees poloidal) detector are generally consistent with the expected first-orbit loss of D- 3 He alpha particle fusion products, with an inferred global reaction rate up to ∼10 16 reactions/sec. A qualitatively similar but unexpectedly large loss occurs 45 degrees poloidally below the outer midplane. This additional loss might be due to ICRH tail ions or to ICRH wave-induced loss of previously confined fusion products

Enhanced loss of fusion products during mode conversion heating in TFTR

International Nuclear Information System (INIS)

Darrow, D.S.; Majeski, R.; Fisch, N.J.; Heeter, R.F.; Herrmann, H.W.; Herrmann, M.C.; Zarnstorff, M.C.; Zweben, S.J.

1995-07-01

Ion Bernstein waves (IBWS) have been generated by mode conversion of ion cyclotron range of frequency (ICRF) fast waves in TFTR. The loss rate of fusion products in these discharges can be large, up to 10 times the first orbit loss rate. The losses are observed at the passing/trapped boundary, indicating that passing particles are being moved onto loss orbits either by increase of their v perpendicular due to the wave, by outward transport in minor radius, or both. The lost particles appear to be DD fusion produced tritons heated to ∼1.5 times their birth energy

Flaking of co-deposited hydrogenated carbon layers on the TFTR limiter

International Nuclear Information System (INIS)

Skinner, C.H.; Gentile, C.A.; Menon, M.M.; Barry, R.E.

1999-01-01

Flaking of co-deposited layers on the inner limiter tiles was recently observed in TFTR. This phenomenon was unexpected and has occurred since the termination of plasma operations on 4 April 1997. Flaking affects approximately 15% of the observable tiles and appears on isotropic graphite but not on carbon fibre composite tiles. Photographic images of the flakes and precise measurements of the limiter geometry are reported. The mobilizability of tritium retained in co-deposited layers is an important factor in safety analyses of future DT reactors. A programme to analyse the flakes and tiles is underway. (author). Letter-to-the-editor

Multichord time-resolved electron temperature measurements by the x-ray absorber-foil method on TFTR

International Nuclear Information System (INIS)

Kiraly, J.; Bitter, M.; Efthimion, P.

1985-09-01

Absorber foils have been installed in the TFTR X-Ray Imaging System to permit measurement of the electron temperature along 10 to 30 chords spaced at 5-12.5 cm with a time resolution of less than 100 μs. The technique uses the ratio of x-ray fluxes transmitted through two different foils. The ratio depends mainly on electron temperature. Simulations show that strong impurity line radiation can distort this ratio. To correct for these effects, special beryllium-scandium filters are employed to select the line-free region between 2 and 4.5 keV. Other filter pairs allow corrections for Fe L and Ni L line radiation as well as Ti K and Ni K emission. Good accuracy is also obtained with simple beryllium filters, provided that impurity corrections are incorporated in the analysis, taking line intensities from the x-ray pulse-height analysis diagnostic. A description of modeling calculations and a comparison of temperature values from this diagnostic with data from the x-ray pulse height analysis, the electron cyclotron emission, and the Thomson scattering diagnostics are presented. Several applications of the absorber foil electron temperature diagnostic on TFTR are discussed

Multichord time-resolved electron temperature measurements by the x-ray absorber-foil method on TFTR

Energy Technology Data Exchange (ETDEWEB)

Kiraly, J.; Bitter, M.; Efthimion, P.; von Goeler, S.; Grek, B.; Hill, K.W.; Johnson, D.; McGuire, K.; Sauthoff, N.; Sesnic, S.

1985-09-01

Absorber foils have been installed in the TFTR X-Ray Imaging System to permit measurement of the electron temperature along 10 to 30 chords spaced at 5-12.5 cm with a time resolution of less than 100 ..mu..s. The technique uses the ratio of x-ray fluxes transmitted through two different foils. The ratio depends mainly on electron temperature. Simulations show that strong impurity line radiation can distort this ratio. To correct for these effects, special beryllium-scandium filters are employed to select the line-free region between 2 and 4.5 keV. Other filter pairs allow corrections for Fe L and Ni L line radiation as well as Ti K and Ni K emission. Good accuracy is also obtained with simple beryllium filters, provided that impurity corrections are incorporated in the analysis, taking line intensities from the x-ray pulse-height analysis diagnostic. A description of modeling calculations and a comparison of temperature values from this diagnostic with data from the x-ray pulse height analysis, the electron cyclotron emission, and the Thomson scattering diagnostics are presented. Several applications of the absorber foil electron temperature diagnostic on TFTR are discussed.

Development of remote welding equipment and techniques for the TFTR vacuum vessel

International Nuclear Information System (INIS)

Masson, L.S.; Watts, K.D.; Larson, R.A.; Aldrich, W.C.

1980-01-01

In the event that the TFTR vacuum vessel is damaged or one of the toroidal field coils fails after the system has become substantially activated, it is necessary to remotely remove and replace the damaged section of the vessel using remote handling procedures. This paper describes a welding system developed through the final design stage to perform the remote welding necessary during the replacement operation. Information is presented describing the vessel configuration, the replacement sequence, the welding system requirements, welder configuration, supporting systems, the weld development program and future development requirements

Model of alpha particle diffusion in the outer limiter shadow of TFTR

International Nuclear Information System (INIS)

Wang, S.; Academia Sinica, Hefei, Anhui; Zweben, S.J.

1996-05-01

A new code, Monte Carlo Collisional Stochastic Orbit Retracing (MCCSOR), has been developed to model the alpha particle loss signal as measured by the outer midplane scintillator detector in TFTR. The shadowing effects due to the outer limiters and the detector itself have been included, along with a pitch angle scattering and stochastic ripple diffusion. Shadowing by the outer limiters has a strong effect on both the magnitude and pitch angle distribution of the calculated loss. There is at least qualitative agreement between the calculated results and the experimental data

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

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

TFTR Ion Cyclotron Range of Frequencies (ICRF) experimental data analysis collaboration. Annual progress report, December 1, 1993--November 30, 1994

International Nuclear Information System (INIS)

Sharer, J.E.; Bettenhausen, M.; Lam, N.; Sund, R.

1994-08-01

The research performed under this grant during the past year has concentrated on coupling, heating, and current drive issues for TFTR. The authors have developed a code and submitted for publication a open-quotes 3-Dclose quotes coupling analysis of the TFIR ICRF cavity-backed coil antennas to plasma edge profiles including the Faraday shield blade angle and fast wave coupling for heating and current drive. They have also carried out TFTR ICRF full-wave field solutions and heating analyses for the second harmonic tritium supershot, and the effects of fusion alpha particle and tritium ion tail populations on the ICRF absorption. They have also published a paper on the effects of alpha particle absorption on fundamental deuterium ion cyclotron absorption incorporating self-consistent deuterium tails and fusion reactivity. Research progress, publications, and conference presentations are summarized in this report

Annual report, October 1, 1978-September 30, 1979

Energy Technology Data Exchange (ETDEWEB)

1981-03-01

This report covers the following areas: (1) PLT, (2) PDX, (3) smaller devices, (4) TFTR, (5) TFTR blanket module experiments, (6) engineering, (7) machine design and fabrication, and (8) advanced projects design and analysis, (9) design studies for new devices, (10) theory, (11) administration, and (12) graduate education. (MOW)

Annual report, October 1, 1978-September 30, 1979

International Nuclear Information System (INIS)

1981-03-01

This report covers the following areas: (1) PLT, (2) PDX, (3) smaller devices, (4) TFTR, (5) TFTR blanket module experiments, (6) engineering, (7) machine design and fabrication, and (8) advanced projects design and analysis, (9) design studies for new devices, (10) theory, (11) administration, and (12) graduate education

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

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

Some features of the disruption instability in reversed shear TFTR plasmas

International Nuclear Information System (INIS)

Semenov, I.B.; Mirnov, S.V.; McGuire, K.M.

1999-01-01

The behaviour of MHD perturbations before and during disruptions in TFTR reversed shear plasmas with q min ∼ 2 was analysed. In the q min region, tearing modes, wavelike modes, and mixed tearing plus wavelike modes are followed by disruption. Sometimes a helical snake (helix) appears at the X point of the q min island. The local outward electron energy transport near the X point can be explained by the development of 'positive' magnetic islands (islands with positive current perturbations). It is proposed that the disruption is initiated when the X point of the magnetic islands coincides in one toroidal position near the torus equator. (author)

The high density and high βpol disruption mechanism on TFTR

International Nuclear Information System (INIS)

Fredrickson, E.D.; Manickam, J.; McGuire, K.M.; Monticello, D.; Nagayama, Y.; Park, W.; Taylor, G.

1992-01-01

Studies of disruptions on TFTR have been extended to include high density disruptions as well as the high β pol disruptions. The data strongly suggests that the (m,n)=(1,1) mode plays an important role in both types of disruptions. Further, for the first time, it is unambiguously shown, using a fast electron cyclotron emission (ECE) instrument for the electron temperature profile measurements, that the (m,n)=(1,1) precursor to the high density disruptions has a 'cold bubble' structure. The precursor to the major disruption at high density resembles the 'vacuum bubble' model of disruptions first proposed by Kadomtsev and Pogutse. (author) 2 refs., 2 figs

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

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

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

Deposition of lithium on a plasma edge probe in TFTR -- Behavior of lithium-painted walls interacting with edge plasmas

Energy Technology Data Exchange (ETDEWEB)

Hirooka, Y. [Univ. of California, San Diego, La Jolla, CA (United States); Ashida, K. [Toyama Univ. (Japan); Kugel, H. [Princeton Univ., NJ (United States)] [and others

1998-05-01

Recent observations have indicated that lithium pellet injection wall conditioning plays an important role in achieving the enhanced supershot regime in TFTR. However, little is understood about the behavior of lithium-coated limiter walls, interacting with edge plasmas. In the final campaign of TFTR, a cylindrical carbon fiber composite probe was inserted into the boundary plasma region and exposed to ohmically-heated deuterium discharges with lithium pellet injection. The ion-drift side probe surface exhibits a sign of codeposition of lithium, carbon, oxygen, and deuterium, whereas the electron side essentially indicates high-temperature erosion. It is found that lithium is incorporated in these codeposits in the form of oxide at the concentration of a few percent. In the electron side, lithium has been found to penetrate deeply into the probe material, presumably via rapid diffusion through interplane spaces in the graphite crystalline. Though it is not conclusive, materials mixing in the carbon and lithium system appears to be a key process in successful lithium wall conditioning.

Three-dimensional Monte Carlo calculations of the neutron and γ-ray fluences in the TFTR diagnostic basement and comparisons with measurements

International Nuclear Information System (INIS)

Liew, S.L.; Ku, L.P.; Kolibal, J.G.

1985-10-01

Realistic calculations of the neutron and γ-ray fluences in the TFTR diagnostic basement have been carried out with three-dimensional Monte Carlo models. Comparisons with measurements show that the results are well within the experimental uncertainties

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

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/)

Annual report, October 1, 1979-September 30, 1980

International Nuclear Information System (INIS)

1981-05-01

This annual report covers research progress on each of the following areas: (1) PLT device, (2) PDX, (3) spheromak, (4) smaller devices, (5) theory, (6) TFTR, (7) applied physics, (8) TFTR blanket module experiments, (9) advanced toroidal facility, (10) advanced projects design and analysis, (11) engineering, and (12) fabrication, operations and maintenance

Annual report, October 1, 1979-September 30, 1980

Energy Technology Data Exchange (ETDEWEB)

1981-05-01

This annual report covers research progress on each of the following areas: (1) PLT device, (2) PDX, (3) spheromak, (4) smaller devices, (5) theory, (6) TFTR, (7) applied physics, (8) TFTR blanket module experiments, (9) advanced toroidal facility, (10) advanced projects design and analysis, (11) engineering, and (12) fabrication, operations and maintenance. (MOW)

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

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.)

QA support for TFTR reliability improvement program in preparation for DT operation

International Nuclear Information System (INIS)

Parsells, R.F.; Howard, H.P.

1987-01-01

As TFTR approaches experiments in the Q=1 regime, machine reliability becomes a major variable in achieving experimental objectives. This paper describes the methods used to quantify current reliability levels, levels required for D-T operations, proposed methods for reliability growth and improvement, and tracking of reliability performance in that growth. Included in this scope are data collection techniques and short comings, bounding current reliability on the upper end, and requirements for D-T operations. Problem characterization through Pareto diagrams provides insight into recurrent failure modes and the use of Duane plots for charting of reliability changes both cumulative and instantaneous, is explained and demonstrated

Measurements of DT alpha particle loss near the outer midplane of TFTR

International Nuclear Information System (INIS)

Zweben, S.J.; Darrow, D.S.; Herrmann, H.W.; Redi, M.H.; Schivell, J.; White, R.B.

1995-07-01

Measurements of DT alpha particle loss to the outer midplane region of TFTR have been made using a radially movable scintillator detector. The conclusion from this data is that mechanisms determining the DT alpha loss to the outer midplane are not substantially different from those for DD fusion products. Some of these results are compared with a simplified theoretical model for TF ripple-induced alpha loss, which is expected to be the dominant classical alpha loss mechanism near the outer midplane. An example of plasma-driven MHD-induced alpha particle loss is shown, but no signs of any ''collective'' alpha instability-induced alpha loss have yet been observed

Vacuum system design and tritium inventory for the TFTR charge exchange diagnostic

International Nuclear Information System (INIS)

Medley, S.S.

1979-05-01

The charge exchange diagnostic for the TFTR is comprised of two analyzer systems which contain a total of twenty independent mass/energy analyzers and one diagnostic neutral beam tentatively rated at 80 keV, 15 A. The associated vacuum systems were analyzed using the Vacuum System Transient Simulator (VSTS) computer program which models the transient transport of multi-gas species through complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. In addition to providing improved design performance at reduced cost, the analysis yields estimates for the exchange of tritium from the torus to the diagnostic components and of the diagnostic working gases to the torus

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

Enhanced loss of fast ions during mode conversion ion Bernstein wave heating in TFTR

International Nuclear Information System (INIS)

Darrow, D.S.; Majeski, R.; Fisch, N.J.; Heeter, R.F.; Herrmann, H.W.; Herrmann, M.C.; Zarnstorff, M.C.; Zweben, S.J.

1995-12-01

A strong interaction of fast ions with ion Bernstein waves has been observed in TFTR. It results in a large increase in the fast ion loss rate, and heats the lost particles to several MeV. The lost ions are observed at the passing/trapped boundary and appear to be either DD fusion produced tritons or accelerated D neutral beam ions. Under some conditions, enhanced loss of DT alpha particles is also seen. The losses provide experimental support for some of the elements required for alpha energy channeling

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

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.)

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

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

Observations of neutral beam and ICRF tail ion losses due to Alfven modes in TFTR

International Nuclear Information System (INIS)

Darrow, D.S.; Zweben, S.J.; Chang, Z.

1996-04-01

Fast ion losses resulting from MHD modes at the Alfven frequency, such as the TAE, have been observed in TFTR. The modes have been driven both by neutral beam ions, at low B T , and by H-minority ICRF tail ions at higher B T . The measurements indicate that the loss rate varies linearly with the mode amplitude, and that the fast ion losses during the mode activity can be significant, e.g. up to 10% of the input power is lost in the worst case

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

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

Use of the TFTR prototype charge exchange neutral analyzer for fast He3++ diagnostics during ICRF heating on PLT

International Nuclear Information System (INIS)

Medley, S.S.

1981-07-01

The Charge Exchange Neutral Analyzer (CENA) for TFTR is designed to measure singly charged ion species of atomic mass A = 1, 2, and 3 simultaneously with up to 75 energy channels per mass and an energy range of 0.5 3 charge exchange neutrals makes the analyzer of particular interest for recently proposed fast He 3 ++ diagnostics during ICRF heating on PLT

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

Anomalous delayed loss of trapped D-D fusion products in TFTR

International Nuclear Information System (INIS)

Zweben, S.J.; Darrow, D.S.; Fredrickson, E.D.; Mynick, H.E.

1993-02-01

A new anomalous delayed loss of D-D fusion products has been measured at the bottom of the TFRR vessel. This loss is delayed by ∼ 0.2 sec with respect to the usual prompt first-orbit loss, and has a correspondingly lower energy, i.e. about half the fusion product birth energy. This loss process dominates the total fusion product loss measured 90 degrees below the midplane for plasma currents. I≥ 1.8 MA and major radii near R=2.45 m, e.g. for recent TFTR supershots. This delayed feature can occur without large coherent MED activity, although it can be strongly modulated by such activity. Several possible causes for this phenomenon are discussed, but no clear explanation for this delayed loss has yet been found

Design and operation of the pellet charge exchange diagnostic for measurement of energetic confined alphas and tritons on TFTR

International Nuclear Information System (INIS)

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

1996-05-01

Radially-resolved energy and density distributions of the energetic confined alpha particles in D-T experiments on TFTR are being measured by active neutral particle analysis using low-Z impurity pellet injection. When injected into a high temperature plasma, an impurity pellet (e.g. Lithium or Boron) rapidly ablates forming an elongated cloud which is aligned with the magnetic field and moves with the pellet. This ablation cloud provides a dense target with which the alpha particles produced in D-T fusion reactions can charge exchange. A small fraction of the alpha particles incident on the pellet ablation cloud will be converted to helium neutrals whose energy is essentially unchanged by the charge transfer process. By measuring the resultant helium neutrals escaping from the plasma using a mass and energy resolving charge exchange analyzer, this technique offers a direct measurement of the energy distribution of the incident high-energy alpha particles. Other energetic ion species can be detected as well, such as tritons generated in D-D plasmas and H or He 3 RF-driven minority ion tails. The diagnostic technique and its application on TFTR are described in detail

Implementation of the α-CHERS diagnostic for D-T operation of TFTR

International Nuclear Information System (INIS)

McKee, G.R.; Fonck, R.J.; Stratton, F.K.

1995-01-01

The α-CHERS diagnostic is a high throughput charge exchange recombination spectroscopy diagnostic designed to measure the density profile and time evolution of 0-500 keV alpha particles during D-T operation of TFTR. Following successful tests with a prototype (α-CHERS system, an improved, multi-channel system has been installed for D-T Operation. Three spatial channels may be observed simultaneously, and the spectral resolution of 0.5 nm permits increased alpha energy resolution and improved impurity line identification. More efficient coupling optics between the spectrometer and CCD detectors have increased the light throughput, and radiation shielding has been installed around the detectors and spectrometers to eliminate the neutron/gamma ray noise observed in high power D-D plasmas

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

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

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

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

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.

Design of the calorimeter and beam dump for the TFTR prototype neutral beam injector

International Nuclear Information System (INIS)

Stone, R.R.; Haughian, J.M.

1977-01-01

A calorimeter has been designed for use with the TFTR prototype neutral beam injection system. It consists of three vees each having two 18.8-mm-thick (0.75 in.) copper plates at a 6-deg angle, relative to the beam centerline. The maximum power density on a plate with this arrangement will be 2.0 kW/cm 2 , resulting in a front surface temperature rise of about 420 0 C. A support and retraction system moves the calorimeter in and out of the beam centerline. Various factors used in the selection of the absorber plate material will be discussed and also some experimental test results will be presented

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

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

Localized MHD activity near transport barriers in JT-60U and TFTR

International Nuclear Information System (INIS)

Manickam, J.

2001-01-01

Localized MHD activity observed in JT-60U and TFTR near transport barriers with their associated large pressure gradients is investigated. Stability analysis of equilibria modeling the experiments supports an identification of this MHD as being due to an ideal MHD n=1 instability. The appearance of the instability depends on the local pressure gradient, local shear in the q profile and the proximity of rational surfaces where q∼m/n and m and n are the poloidal and toroidal mode numbers respectively. The mode width is shown to depend on the local value of q, and is larger when q is smaller. In addition the role of the edge current density in coupling the internal mode to the plasma edge and of the energetic particles which can drive fishbone like modes is investigated. (author)

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

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

Engineering study of the neutral beam and rf heating systems for DIII-D, MFTF-B, JET, JT-60 and TFTR

International Nuclear Information System (INIS)

Lindquist, W.B.; Staten, S.H.

1987-01-01

An engineering study was performed on the rf and neutral beam heating systems implemented for DIII-D, MFTF-B, JET, JT-60 and TFTR. Areas covered include: methodology used to implement the systems, technology, cost, schedule, performance, problems encountered and lessons learned. Systems are compared and contrasted in the areas studied. Summary statements were made on common problems and lessons learned. 3 refs., 6 tabs

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.

Structural analysis of TFTR vacuum vessel bellows and bellows cover sections

International Nuclear Information System (INIS)

Driesen, G.

1975-10-01

A structural evaluation of the bellows and bellows cover sections was undertaken in order to confirm the structural integrity of these TFTR vacuum vessel components in the prescribed operating environment. The evaluations investigate component stability, stress, and deflection behavior. The products of this investigation appearing in this report include; (1) Structural verification of the vacuum bellows as currently defined in an operating environment of one atmosphere external pressure and 93 0 C (200 0 F) uniform temperature. (2) The establishment of a structurally adequate design configuration for the bellows cover section. (3) The presentation of a parametric study which indicates the effects of varying some bellows cover section parameters in order to obtain acceptable variations of this design configuration. (4) A verification of bellows and bellows cover section integrity to preclude a fatigue type failure for reactor startup and shutdown cyclic life in the design environment

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

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

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

Measurements of confined alphas and tritons in the MHD quiescent core of TFTR plasmas using the pellet charge exchange diagnostic

International Nuclear Information System (INIS)

Medley, S.S.; Budny, R.V.; Mansfield, D.K.

1996-05-01

The energy distributions and radial density profiles of the fast confined trapped alpha particles in DT experiments on TFTR are being measured in the energy range 0.5--3.5 MeV using a Pellet Charge eXchange (PCX) diagnostic. A brief description of the measurement technique which involves active neutral particle analysis using the ablation cloud surrounding an injected impurity pellet as the neutralizer is presented. This paper focuses on alpha and triton measurements in the core of MHD quiescent TFTR discharges where the expected classical slowing down and pitch angle scattering effects are not complicated by stochastic ripple diffusion and sawtooth activity. In particular, the first measurement of the alpha slowing down distribution up to the birth energy, obtained using boron pellet injection, is presented. The measurements are compared with predictions using either the TRANSP Monte-Carlo code and/or a Fokker-Planck Post-TRANSP processor code, which assumes that the alphas and tritons are well confined and slow down classically. Both the shape of the measured alpha and triton energy distributions and their density ratios are in good agreement with the code calculations. The authors conclude that the PCX measurements are consistent with classical thermalization of the fusion-generated alphas and tritons

Simulations of alpha parameters in a TFTR DT supershot with high fusion power

International Nuclear Information System (INIS)

Budny, R.V.; Bell, M.G.; Janos, A.C.

1995-07-01

A TFTR supershot with a plasma current of 2.5 MA, neutral beam heating power of 33.7 MW, and a peak DT fusion power of 7.5 MW is studied using the TRANSP plasma analysis code. Simulations of alpha parameters such as the alpha heating, pressure, and distributions in energy and v parallel /v are given. The effects of toroidal ripple and mixing of the fast alpha particles during the sawteeth observed after the neutral beam injection phase are modeled. The distributions of alpha particles on the outer midplane are peaked near forward and backward v parallel /v. Ripple losses deplete the distributions in the vicinity of v parallel /v ∼-0.4. Sawtooth mixing of fast alpha particles is computed to reduce their central density and broaden their width in energy

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

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

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

Unifying role of radial electric field shear in the confinement trends of transitionless regimes in TFTR

International Nuclear Information System (INIS)

Ernst, D.R.; Beer, M.; Batha, S.

2001-01-01

Turbulence suppression by radial electric field shear (E r ) is shown to be important in the enhanced confinement of TFTR supershot plasmas. Simulations of supershot ion temperature profiles are performed using an existing parameterization of transport due to toroidal ion temperature gradient modes, extended to include suppression by E r shear. New spectroscopic measurements of E r differ significantly from prior neoclassical estimates. Supershot temperature profiles appear to be consistent with a criterion describing near-complete turbulence suppression by intrinsically generated E r shear. Helium spoiling and xenon puffing experiments are simulated to illustrate the role of E r shear in the confinement changes observed. (author)

Unifying role of radial electric field shear in the confinement trends of transitionless regimes in TFTR

International Nuclear Information System (INIS)

Ernst, D.R.; Beer, M.; Batha, S.

1999-01-01

Turbulence suppression by radial electric field shear (E r ) is shown to be important in the enhanced confinement of TFTR supershot plasmas. Simulations of supershot ion temperature profiles are performed using an existing parameterization of transport due to toroidal ion temperature gradient modes, extended to include suppression by E r shear. New spectroscopic measurements of E r differ significantly from prior neoclassical estimates. Supershot temperature profiles appear to be consistent with a criterion describing near-complete turbulence suppression by intrinsically generated E r shear. Helium spoiling and xenon puffing experiments are simulated to illustrate the role of E r shear in the confinement changes observed. (author)

Recent progress on MHD-induced loss of D-D fusion products in TFTR

International Nuclear Information System (INIS)

Zweben, S.J.; Darrow, D.S.; Budny, R.V.; Cheng, C.Z.; Fredrickson, E.D.; Herrmann, H.; Mynick, H.E.; Schivell, J.

1993-08-01

This paper reviews the recent progress made toward understanding the MHD-induced loss of D-D fusion products which has been seen on TFTR since 1988. These measurements have been made using the ''lost alpha'' diagnostic, which is described briefly. The largest MHD- induced loss occurs with coherent 3/2 or 2/1 MHD activity (kink/tearing modes), which can cause up to ∼3--5 times the first-orbit loss at I∼1.6--1.8 MA, roughly a ∼20--30% global los of D-D fusion products. Modeling of these MHD-induced losses has progressed to the point where the basic loss mechanism can be accounted for qualitatively, but the experimental results can not yet be understood quantitatively. Several alpha loss codes are being developed to improve the quantitative comparison between experiment and theory

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

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

A labview approach to instrumentation for the TFTR bumper limiter alignment project

International Nuclear Information System (INIS)

Skelly, G.N.; Owens, D.K.

1992-01-01

This paper reports on a project recently undertaken to measure the alignment of the TFTR bumper limiter in relation to the toroidal magnetic field axis. The process involved the measurement of the toroidal magnetic field, and the positions of the tiles that make up the bumper limiter. The basis for the instrument control and data acquisition system was National Instrument's LabVIEW 2. LabVIEW is a graphical programming system for developing scientific and engineering applications on a Macintosh. For this project, a Macintosh IIci controlled the IEEE-488 GPIB programmable instruments via an interface box connected to the SCSI port of the computer. With LabVIEW, users create graphical software modules called virtual instruments instead of writing conventional text-based code. To measure the magnetic field, the control system acquired data from two nuclear magnetic resonance magnetometers while the torroidal field coils were pulsed. To measure the position of the tiles on the limiter, an instrumented mechanical arm was used inside the vessel

ICRF antenna modifications and additions for TFTR: Relevance to BPX/ITER projections

International Nuclear Information System (INIS)

Hosea, J.; Phillips, C.K.; Raftopoulos, S.; Stevens, J.; Wilson, J.R.

1991-01-01

The TFTR Bay L and M antennas have been modified to improve their power handling capability. In particular, the Bay L antenna, which exhibited a lower than expected loading resistance, now has a configuration similar to that of Bay M -- slotted walls and septum -- and together with Bay M is expected to support 7 MW operations. The in situ loading enhancement achieved for the Modified Bay L design will serve to quantify models for the coupling effects of slots. Also, comparisons with Bay M loading performance will elucidate wave spectrum and antenna location (relative to in-vessel structures) effects. Two new antennas, with single/double row shields slanted at 6 degree (along B) are to be added in the near future to augment the power capability to ∼12.5 MW. The relevance of the four antenna array features to quantifying BPX/ITER antenna characteristic projections for heating and current drive is discussed. 8 refs., 5 figs

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

The TFTR 40 MW neutral beam injection system and DT operations

International Nuclear Information System (INIS)

Stevenson, T.; O'Connor, T.; Garzotto, V.

1995-01-01

Since December 1993, TFTR has performed DT experiments using tritium fuel provided mainly by neutral beam injection. Significant alpha particle populations and reactor-like conditions have been achieved at the plasma core, and fusion output power has risen to a record 10.7 MW using a record 40 MW NB heating. Tritium neutral beams have injected into over 480 DT plasmas and greater than 500 kCi have been processed through the neutral beam gas, cryo, and vacuum systems. Beam tritium injections, as well as tritium feedstock delivery and disposal, have now become part of routine operations. Shot reliability with tritium is about 90% and is comparable to deuterium shot reliability. This paper describes the neutral beam DT experience including the preparations, modifications, and operating techniques that led to this high level of success, as well as the critical differences in beam operations encountered during DT operations. Also, the neutral beam maintenance and repair history during DT operations, the corrective actions taken, and procedures developed for handling tritium contaminated components are discussed in the context of supporting a continuous DT program

Absolute calibration of TFTR neutron detectors for D-T plasma operation

International Nuclear Information System (INIS)

Jassby, D.L.; Johnson, L.C.; Roquemore, A.L.; Strachan, J.D.; Johnson, D.W.; Medley, S.S.; Young, K.M.

1995-03-01

The two most sensitive TFTR fission-chamber detectors were absolutely calibrated in situ by a D-T neutron generator (∼5 x 10 7 n/s) rotated once around the torus in each direction, with data taken at about 45 positions. The combined uncertainty for determining fusion neutron rates, including the uncertainty in the total neutron generator output (±9%), counting statistics, the effect of coil coolant, detector stability, cross-calibration to the current mode or log Campbell mode and to other fission chambers, and plasma position variation, is about ±13%. The NE-451 (ZnS) scintillators and 4 He proportional counters that view the plasma in up to 10 collimated sightlines were calibrated by scanning. the neutron generator radially and toroidally in the horizontal midplane across the flight tubes of 7 cm diameter. Spatial integration of the detector responses using the calibrated signal per unit chord-integrated neutron emission gives the global neutron source strength with an overall uncertainty of ±14% for the scintillators and ±15% for the 4 He counters

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

Next Step Spherical Torus Design Studies

International Nuclear Information System (INIS)

Neumeyer, C.; Heitzenroeder, P.; Kessel, C.; Ono, M.; Peng, M.; Schmidt, J.; Woolley, R.; Zatz, I.

2002-01-01

Studies are underway to identify and characterize a design point for a Next Step Spherical Torus (NSST) experiment. This would be a ''Proof of Performance'' device which would follow and build upon the successes of the National Spherical Torus Experiment (NSTX) a ''Proof of Principle'' device which has operated at PPPL since 1999. With the Decontamination and Decommissioning (DandD) of the Tokamak Fusion Test Reactor (TFTR) nearly completed, the TFTR test cell and facility will soon be available for a device such as NSST. By utilizing the TFTR test cell, NSST can be constructed for a relatively low cost on a short time scale. In addition, while furthering spherical torus (ST) research, this device could achieve modest fusion power gain for short-pulse lengths, a significant step toward future large burning plasma devices now under discussion in the fusion community. The selected design point is Q=2 at HH=1.4, P subscript ''fusion''=60 MW, 5 second pulse, with R subscript ''0''=1.5 m, A=1.6, I subscript ''p''=10vMA, B subscript ''t''=2.6 T, CS flux=16 weber. Most of the research would be conducted in D-D, with a limited D-T campaign during the last years of the program

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

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

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

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

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.

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

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

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

TFTR lithium blanket module program. Final design report. Volume VII. LBM instrumentation and Program of Experiments and Analysis

International Nuclear Information System (INIS)

Harker, Y.D.; Tsang, F.Y.; Jassby, D.L.

1983-07-01

The Program of Experiments and Analysis comprises 3 spheres of activity: (1) measurements of neutron fluence and flux spectra inside and around the LBM, and of tritium production in the LBM central zone; (2) neutronic-code modeling and analysis of the TFTR/LBM system to predict the quantities measured in (1); (3) comparisons of the predicted and measured quantities, and improvements of the code modeling and analysis and the experimental techniques, in order to resolve any discrepancy between prediction and measurement. The measurement techniques are discussed. Section 5 of this volume discusses the strategy for carrying out the measurement program, for making comparisons with the neutronics code predictions, and for resolving discrepancies

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

ETF design activities: status and future direction

International Nuclear Information System (INIS)

Steiner, D.

1980-01-01

The nature of the device to follow TFTR has been evolving over the past six years. The most recent design activities have been directed to the Engineering Test Facility (ETF) concept. In order to address the design of the ETF a centralized laboratory/industry design team was established and hosted at ORNL. This Design Center began ETF Preconceptual design in October, 1979. The ETF design effort has been judged to be sound. However, the ETF mission has been judged to be overly ambitious. It has been recommended that the device to follow TFTR be dedicated to the investigation of engineering feasibility and that device has been designated the Fusion Engineering Device (FED). The ETF design activities provide a basis for developing the FED design. This will be accomplished under the direction of the newly formed FED Technical Management Board

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.

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

Control system implementation for a complex low inventory cryogenic distillation system for Princeton TFTR

International Nuclear Information System (INIS)

Busigin, A.; Busigin, C.J.; Adamek, F.; Woodall, K.B.; Robins, J.R.; Bellamy, D.G.; Fong, C.; Kalyanam, K.M.; Sood, S.K.

1995-01-01

The TFTR Tritium Purification System (TPS) is based on a Pd/Ag diffuser front-end for separating hydrogen isotopes from inert gas, and a four column cryogenic distillation cascade for separation of hydrogen isotopes. The system has a tritium inventory of approximately 0.5 g while successfully producing pure H 2 , D 2 and T 2 products. The system has recently been built and successfully commissioned with protium and deuterium. Stable automatic control of the cascade has been demonstrated even when feed rate and composition varied. The automatic control scheme maintained stable column inventories and excellent H 2 and D 2 product qualities. The control system employed new control concepts such as real time analysis of mid-column composition using temperature and pressure data for feedback control. Very stable column inventory control was achieved by automatic adjustment of inter-column flows (feed forward and feed back). This paper discusses the control system design and presents performance test results. (orig.)

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

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

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

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

Progress of research and development of nuclear fusion and development of large nuclear fusion device technology

International Nuclear Information System (INIS)

1994-01-01

In the last several years, the results of tokamak experiments were conspicuous, and the progress of plasma confinement performance, transport mechanism, divertors and impurities, helium transport and exhaust, electric current drive, magnetic field ripple effect and high speed particle transport and DT experiment are reported. The other confinement methods than tokamak, the related theories and reactor technology are described. The conceptual design of ITER was carried out by the cooperation of Japan, USA, EC and the former USSR. The projects of developing nuclear fusion in various countries, the design and the required research and development of ITER, the reconstruction and the required research and development of JT-60, JET and TFTR, the design and the required research and development of large helical device, the state of research and development of laser nuclear fusion and inversion magnetic field pinch nuclear fusion, the activities and roles of industrial circles in large nuclear fusion device technology, and the long term perspective of the technical development of nuclear fusion are described. (K.I.)

Low-Z coating as a first wall of nuclear fusion devices

International Nuclear Information System (INIS)

Shikama, Tatsuo; Okada, Masatoshi

1984-01-01

The tokamak nuclear fusion devices of the largest scale in the world, TFTR in USA and JET in Europe, started the operation from the end of 1982 to 1983. Also in Japan, the tokamak JT-60 is scheduled to begin the operation in 1985. One of the technological obstacles is the problem of first walls facing directly to plasma and subjected to high particle loading and thermal loading. Moreover, first walls achieve the active role of controlling impurities in plasma and recycling hydrogen isotopes. It is impossible to find a single material which satisfies all these requirements. The compounding of materials can create a material having new function, but also has the meaning of expanding the range of material selection. One of the material compounding methods is surface coating. In this paper, as the materials for first walls, the characteristics of low Z materials are discussed from the design examples of actual takamak nuclear fusion devices. The outline of first walls is explained. High priority is given to the impurity control in plasma, and in view of plasma energy emissivity and the rate of self sputtering, low Z material coating seems to be the solution. The merits and the problems of such low Z material coating are discussed. (Kako, I.)

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

Measurements of DT and DD neutron yields by neutron activation on TFTR

International Nuclear Information System (INIS)

Barnes, C.W.; Larson, A.R.; LeMunyan, G.

1994-01-01

A variety of elemental foils have been activated by neutron fluence from TFTR under conditions with the DT neutron yield per shot ranging from 10 12 to over 10 18 , and with the DT/(DD+DT) neutron ratio varying from 0.5% (from triton burnup) to unity. Linear response over this large dynamic range is obtained by reducing the mass of the foils and increasing the cooling time, all while accepting greatly improved counting statistics. Effects on background gamma-ray lines from foil-capsule-material contaminants. and the resulting lower limits on activation foil mass, have been determined. DT neutron yields from dosimetry standard reactions on aluminum, chromium, iron, nickel, zirconium, and indium are in agreement within the ±9% (one-sigma,) accuracy of the measurements: also agreeing are yields from silicon foils using the ACTL library cross-section. While the ENDF/B-V library has too low a cross-section. Preliminary results from a variety of other threshold reactions are presented. Use of the 115 In(n,n) 115m In reaction (0.42 times as sensitive to DT neutrons as DD neutrons) in conjunction with pure-DT reactions allows a determination of the DT/(DD+DT) ratio in trace tritium or low-power tritium beam experiments

Measurements of DT and DD neutron yields by neutron activation on TFTR

International Nuclear Information System (INIS)

Barnes, C.W.; Larson, A.R.; LeMunyan, G.

1995-03-01

A variety of elemental foils have been activated by neutron fluence from TFTR under conditions with the DT neutron yield per shot ranging from 10 12 to over 10 18 , and with the DT/(DD+DT) neutron ratio varying from 0.5% (from triton burnup) to unity. Linear response over this large dynamic range is obtained by reducing the mass of the foils and increasing the cooling time, all while accepting greatly improved counting statistics. Effects on background gamma-ray lines from foil-capsule-material contaminants, and the resulting lower limits on activation foil mass, have been determined. DT neutron yields from dosimetry standard reactions on aluminum, chromium, iron, nickel, zirconium, and indium are in agreement within the ±9% (one-sigma) accuracy of the measurements; also agreeing are yields from silicon foils using the ACTL library cross-section, while the ENDF/B-V library has too low a cross-section. Preliminary results from a variety of other threshold reactions are presented. Use of the 115 In(n.n') 115m In reaction (0.42 times as sensitive to DT neutrons as DD neutrons) in conjunction with pure-DT reactions allows a determination of the DT/(DD+DT) ratio in trace tritium or low-power tritium beam experiments

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

Installation and early operation of a complex low inventory cryogenic distillation system for the Princeton TFTR

International Nuclear Information System (INIS)

Busigin, A.; Busigin, C.J.; Robins, J.R.; Woodall, K.B.; Bellamy, D.G.; Fong, C.; Kalyanam, K.; Sood, S.K.

1995-01-01

A low inventory Tritium Purification System (TPS) has just been installed at the Princeton Plasma Physics Laboratory (PPPL). The TPS was designed specifically for PPPL, based on their specifications for exhaust gases. The generic design, however, can easily be modified to accept a large variety of input conditions. The Princeton system is designed to have a total tritium inventory of approximately 0.5 g while producing pure product streams consisting of H 2 , D 2 , and T 2 . The purpose of the system is to separate and recycle unburnt tritium from the TFTR and to produce hydrogen and deuterium streams that are free of tritium. These streams can be disposed by stacking, thus eliminating the need to create large volume waste streams that are contaminated with tritium and that must be managed for permanent disposal. This paper will discuss the installation, the modifications and preliminary results of operation of this system at Princeton. 2 refs., 3 figs

Study of the mass and energy resolution of the E parallel B charge exchange analyzer for TFTR

International Nuclear Information System (INIS)

Kaita, R.; Medley, S.S.

1979-09-01

The charge exchange diagnostic for TFTR requires simultaneous multispecie (H + , D + , γ + ) analysis of particles in the energy range of 0.5 - 150 keV. The analyzer design chosen to provide this capability employs a wide gap semi-circular region of superimposed parallel electric and magnetic fields to accomplish mass and energy resolution, respectively. Combined with a large area, multi-anode microchannel plate detector, this arrangement will enable the energy distributions of protons, deuterions, and tritons to be measured concurrently as a function of time during each discharge. A computer simulation program for calculating ion trajectories through the analyzer was written that includes a realistic model of the magnetic and electrostatic fringe fields. This report presents the results of a study of the proposed E parallel B analyzer, and it reveals that the fringe fields are not detrimental to the performance of the analyzer

Confined trapped-alpha behavior in TFTR deuterium-tritium plasmas

International Nuclear Information System (INIS)

Medley, S.S.; Budny, R.V.; Redi, M.H.; Roquemore, A.L.; White, R.B.; Petrov, M.P.; Gorelenkov, N.N.

1997-10-01

Confined trapped-alpha energy spectra and differential radial density profiles in TFTR D-T plasmas are obtained with the Pellet Charge-eXchange (PCX) diagnostic which measures high energy (E α = 0.5--3.5 MeV), trapped alphas (v parallel /v = - 0.048) at a single time slice (Δt ∼ 1 msec) with a spatial resolution of Δr ∼ 5 cm. Tritons produced in D-D plasmas and RF-driven ion tails (H, 3 He or T) were also observed and energetic tritium ion tail measurements will be discussed. PCX alpha and triton energy spectra extending up to their birth energies were measured in the core of MHD-quiescent discharges where the expected classical slowing down and pitch angle scattering effects are not complicated by stochastic ripple diffusion and sawtooth activity. Both the shape of the measured alpha and triton energy distributions and their density ratios are in good agreement with TRANSP predictions, indicating that the PCX measurements are consistent with classical thermalization of the fusion-generated alphas and tritons. From calculations, these results set an upper limit on possible anomalous radial diffusion for trapped alphas of D α ≤ 0.01 m 2 s -1 . Outside the core, where the trapped alphas are influenced by stochastic ripple diffusion effects, the PCX measurements are consistent with the functional dependence of the Goldston-White-Boozer stochastic ripple threshold on the alpha energy and the q-profile. In the presence of strong sawtooth activity, the PCX diagnostic observes significant redistribution of the alpha signal radial profile wherein alphas are depleted in the core and redistributed to well outside the q = 1 radius, but apparently not beyond the energy-dependent stochastic ripple loss boundary

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

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

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

The relationship between turbulence measurements and transport in different heating regimes in TFTR

International Nuclear Information System (INIS)

Bretz, N.L.; Mazzucato, E.; Nazikian, R.; Paul, S.F.; Hammett, G.; Rewoldt, G.; Tang, W.M.; Zarnstorff, M.C.

1992-01-01

The scaling of broad band density fluctuations in the confinement zone of TFTR measured by microwave scattering, beam emission spectroscopy (BES), and reflectometry show a relationship between these fluctuations and energy transport measured from power balance calculations. In L-mode plasmas scattering and BES indicates that the density fluctuation level, δn 2 , in the confinement zone for 0.2 aux and I p in a way that is consistent with variations in energy transport. Fluctuation levels measured with all systems increase strongly toward the edge in all heating regimes following increases in energy transport coefficients. Measurements using BES have shown that poloidal and radial correlation lengths in the confinement zone of L-mode and supershot plasmas fall in the range of 1 to 2 cm. with a wave structure which has k max ∼ 1 cm -1 (k perpendicular ps ∼ 0.2) in the poloidal direction and k max approaching zero in the radial direction. A simple estimate of the diffusion coefficient based on a measured radial correlation length and correlation time indicates good agreement with power balance calculations. Similar estimates using reflectometry give radial coherence lengths at 10 to 20 kHz in low density ohmic and supershot plasmas of between I and 2 cm

Design of TFTR movable limiter blades for ohmic and neutral-beam-heated plasmas

International Nuclear Information System (INIS)

Doll, D.W.; Ulrickson, M.A.; Cecchi, J.L.; Citrolo, J.C.; Weissenburger, D.; Bialek, J.

1981-10-01

A new set of movable limiter blades has been designed for TFTR that will meet both the requirements of the 4 MW ohmic heated and the 33 MW neutral beam heated plasmas. This is accomplished with three limiter blades each having and elliptical shape along the toroidal direction. Heat flux levels are acceptable for both ohmic heated and pre-strong compression plasmas. The construction consists of graphite tiles attached to cooled backing plates. The tiles have an average thickness of approx. 4.7 cm and are drawn against the backing plate with spring loaded fasteners that are keyed into the graphite. The cooled backing plate provides the structure for resisting disruption and fault induced loads. A set of rollers attached to the top and bottom blades allow them to be expanded and closed in order to vary the plasma surface for scaling experiments. Water cooling lines penetrate only the mid-plane port cover/support plate in such a way as to avoid bolted water connections inside the vacuum boundary and at the same time allow blade movement. Both the upper and lower blades are attached to the mid-plane limiter blade through pivots. Pivot connections are protected against arcing with an alumina coating and a shunt bar strap. Remote handling is considered throughout the design

Alpha heating in toroidal devices

International Nuclear Information System (INIS)

Miley, G.H.

1978-01-01

Ignition (or near-ignition) by alpha heating is a key objective for the achievement of economic fusion reactors. While good confinement of high-energy alphas appears possible in larger reactors, near-term tokamak-type ignition experiments as well as some concepts for small reactors (e.g., the Field-Reversed Mirror or FRM) potentially face marginal situations. Consequently, there is a strong motivation to develop methods to evaluate alpha losses and heating profiles in some detail. Such studies for a TFTR-size tokamak and for a small FRM are described here

Neutron and hard x-ray measurements during pellet deposition in TFTR

Energy Technology Data Exchange (ETDEWEB)

Heidbrink, W.W.; Milora, S.L.; Schmidt, G.L.; Schneider, W.; Ramsey, A.

1986-06-01

Measurements of neutrons and hard x rays are made with a pair of plastic scintillators during injection of deuterium pellets into deuterium TFTR plasmas. Three cases are investigated. During ohmic heating in plasmas with few runaway electrons, the neutron emission does not increase when a pellet is injected, indicating that strong acceleration of the pellet ions does not occur. In ohmic plasmas with low but detectable levels of runaway electrons, an x-ray burst is observed on a detector near the pellet injector as the pellet ablates, while a detector displaced 126/sup 0/ toroidally from the injector does not measure a synchronous burst. Reduced pellet penetration correlates with the presence of x-ray emission, suggesting that the origin of the burst is bremsstrahlung from runaway electrons that strike the solid pellet. In deuterium beam-heated discharges, an increase in the d-d neutron emission is observed when the pellet ablates. In this case, the increase is due to fusion reactions between beam ions and the high density neutral and plasma cloud produced by ablation of the pellet; this localized density perturbation equilibrates in about 700 ..mu..sec. Analysis of the data indicates that the density propagates without forming a sharp shock front with a rapid initial propagation velocity (greater than or equal to 2 x 10/sup 7/ cm/sec) that subsequently decreases to around 3 x 10/sup 6/ cm/sec. Modelling suggests that the electron heat flux into the pellet cloud is much less than the classical Spitzer value.

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

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.

Difference in electron thermal diffusivity and profile between interior and exterior of TFTR L-mode plasmas

International Nuclear Information System (INIS)

Hiroe, S.; Johnson, D.W.; Goldston, R.J.

1990-01-01

The local properties such as scale lengths of the electron density (L n e ), temperature (L T e ), and pressure (L p e ), and the electron thermal diffusivity χ e (r) (m 2 /s) for r/a > 0.3 have been studied for TFTR L-mode discharges under the assumption of χ e = χ i . The scale lengths and the electron thermal diffusivity in the interior 0.3 e can be expressed as (with correlation coefficient R = 0.61), χ e (r) = 1.44 x 10 18 (r/a) 1.0 T e (r) 0.1 q(r) 0.1 /n e 0.9 (r). In the exterior region (0.55 e can be described as (with R = 0.68), χ e (r) = 2.3 x 10 3 (r/a) 1.7 T e (r) 0.7 q(r) 0.8 /n e 0.2 (r). It is interesting to note the negative n e dependence of χ e in the interior and the positive T e dependence of χ e in the exterior

The configuration development and integration of the TPX device

International Nuclear Information System (INIS)

Brown, T.G.

1993-01-01

The TPX configuration was designed to meet the physics objectives and subsystem requirements in an arrangement that allowed access for remote maintenance. The steady state operations of TPX favored the use of superconducting magnets for both the toroidal and poloidal field systems. The desire to react the TF centering and overturning forces in a simplified wedged system lead to a TF case concept incorporating ''two-coil'' TF modules in a 90 degrees four-coil quadrant arrangement. Low ripple and tangential plasma access to accommodate TFTR neutral beams were leading factors in determining the size and number of TF coils. The need for a large amount of space for the divertor and first wall component coolant services further influenced the shaping of the vacuum vessel. Additional configuration influences included: low activation considerations, diverter pumping, remote maintenance requirements, service access and compatibility with the existing TFTR text cell facility. The TPX configuration development and integration process has evolved through the conceptual design period and is now ready to enter the Preliminary Design Phase of the project. This paper describes the status of the configuration development and integration of the major TPX tokamak subsystems components

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

Parameter studies for a two-component fusion experiment

International Nuclear Information System (INIS)

Towner, H.H.

1975-01-01

The sensitivity of the energy multiplication of a two-component fusion experiment is examined relative to the following parameters: energy confinement time (tau/sub E/), particle confinement time (tau/sub p/), effective Z of the plasma (Z/sub eff/), injection rate (j/sub I/) and injection energy (E/sub I/). The Energy Research and Development Administration recently approved funding for such a fusion device (the Toroidal Fusion Test Reactor or TFTR) which will be built at the Princeton Plasma Physics Laboratory. Hence, such a parameter study seems both timely and necessary. This work also serves as an independent check on the design values proposed for the TFTR to enable it to achieve energy breakeven (F = 1). Using the nominal TFTR design parameters and a self-consistent ion-electron power balance, the maximum F-value is found to be approximately 1.2 which occurs at an injection energy of approximately 210 KeV. The injector operation, i.e. its current and energy capability are shown to be a very critical factor in the TFTR performance. However, if the injectors meet the design objectives, there appears to be sufficient latitude in the other parameters to offer reasonable assurance that energy breakeven can be achieved. (U.S.)

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.

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

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

Robotics and remote maintenance concepts for fusion machines

International Nuclear Information System (INIS)

1989-02-01

Descriptions of operation and maintenance of current tokamaks (TFTR, JET, JT-60) is discussed in the context of radioactivation resulting from thermonuclear reactions. Plans for future devices (NET, CIT, FGR) with respect to remote handling, maintenance, measurements, and robotics are discussed. Refs, figs and tabs

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.

Tokamak Fusion Test Reactor - CICADA, an overview

International Nuclear Information System (INIS)

Sauthoff, N.; Bosco, J.; Daniels, R.

1983-01-01

The Central Instrumentation, Control and Data Acquisition (CICADA) System provides the interface between the human operators and the equipment to be controlled and monitored in support of TFTR operation. The functions can be partitioned into those that are instantaneous and fundamental components of a processed control system (such as displaying the current monitored value of a point or changing the state of a piece of equipment) and those that are more complex, being composite actions executed in sequences either related to the TFTR shot cycle or in the execution of timed procedures for the operation of subsystems. In this paper, the authors present the configuration of the equipment ''from the outside in'' - meaning from the man and equipment external interfaces inward toward the components internal to the CICADA system; they next present an overview of the fundamental operations, including the concept of the CICADA devices, the assignment of values to the elements of the definition of the device, the monitoring procedure for all of the devices on the system, the control process including the assignment of control authorization, the handling of special devices such as diagnostic data acquisition and timing modules, and graphics. Next they proceed to discuss composite actions and timed sequences which support the TFTR cycle; included are the system for activating, suspending, and resuming tasks in response to the occurrence of hardware and software events, and the typical sequence of tasks involved in operation of a subsystem for data acquisition during the shot cycle and the archival of raw and results data; special cases will be the real time control systems for the plasma position and current and the gas injection system, both of which provide closed-loop feedback control of plasma position, current and lineintegrated density throughout the shot. Finally, future directions and initiatives discussed

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

Challenges for Plasma Diagnostic in a Next Step Device (FIRE)

International Nuclear Information System (INIS)

Young, Kenneth M.

2002-01-01

The physics program of any next-step tokamak such as FIRE [Fusion Ignition Research Experiment] sets demands for plasma measurement which are at least as comprehensive as on present tokamaks, with the additional capabilities needed for control of the plasma and for understanding the effects of the alpha-particles. The diagnostic instrumentation must be able to provide the fine spatial and temporal resolution required for the advanced tokamak plasma scenarios. It must also be able to overcome the effects of neutron- and gamma-induced electrical noise in ceramic components or detectors, and fluorescence and absorption in optical components. There are practical engineering issues of minimizing radiation streaming while providing essential diagnostic access to the plasma. Many diagnostics will require components at or close to the first wall, e.g., ceramics and MI cable for magnetic diagnostics and mirrors for optical diagnostics; these components must be mounted to operate, and survive, i n fluxes which require special material selection. A better set of diagnostics of alpha-particles than that available for the TFTR [Tokamak Fusion Test Reactor] is essential; it must be qualified well before moving into D-T [deuterim-tritium] experiments. A start has been made to assessing the potential implementation of key diagnostics for the FIRE device. The present status is described

Thermal and hydraulic analyses of TFTR cooling water system and magnetic field coils

International Nuclear Information System (INIS)

Lee, A.Y.

1975-10-01

The TFTR toroidal field coils, ohmic heating, hybrid and equilibrium field coils are cooled by water from the machine area cooling water system. The system has the following major equipment and capacities: flow rate of 3600 gpm; ballast tank volume of 5500 gal; pumps of 70.4 m head; chiller refrigeration rating of 3300 tons and connecting pipe of 45.7 cm I.D. The performance of the closed loop system was analyzed and found to be adequate for the thermal loads. The field coils were analyzed with detailed thermal and hydraulic models, including a simulation of the complete water cooling loop. Under the nominal operating mode of one second of toroidal field flat top time and 300 seconds of pulse cycle time, the maximum temperature for the TF coils is 53 0 C; for the OH coils 46 0 C and for the EF coils 39 0 C, which are well below the coil design limit of 120 0 C. The maximum TF coil coolant temperature is 33 0 C which is below the coolant design limit of 100 0 C. The overall pressure loss of the system is below 6.89 x 10 5 Pa (100 psi). With the given chiller refrigeration capacity, the TF coils can be operated to yield up to 4 seconds of flat top time. The TF coils can be operated on a steady state basis at up to 20% of the pulsed duty design current rating of 7.32 kA/coil

Annual report covering the period October 1, 1980 to September 30, 1981

International Nuclear Information System (INIS)

Phillips, C.A.; Jones, N.D.

1981-01-01

Research during this period is described for each of the following areas: (1) Princeton Large Torus, (2) Poloidal Divertor Experiment, (3) Spheromak, (4) smaller devices, (5) x-ray laser studies, (6) theory, (7) TFTR, (8) applied physics, (9) design studies for new devices, (10) advanced projects design and analysis, (11) engineering, (12) fabrication, operations, and maintenance (13) Projects Office, (14) Computer Division, (15) Administration, and (16) Graduate Education

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

Operation of a TFTR ion source with a ground potential gas feed into the neutralizer

International Nuclear Information System (INIS)

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

1991-01-01

TFTR long pulse ion sources have been operated with gas fed only into the neutralizer. Gas for the plasma generator entered through the accelerator rather than directly into the arc chamber. This modification has been proposed for tritium beam operation to locate control electronics at ground potential and to simplify tritium plumbing. Source operation with this configuration and with the nominal gas system that feeds gas into both the ion source and the center of the neutralizer are compared. Comparison is based upon accelerator grid currents, beam composition, and neutral power delivered to the calorimeter. Charge exchange in the accelerator can be a significant loss mechanism in both systems at high throughput. A suitable operating point with the proposed system was found that requires 30% less gas than used presently. The extracted D + , D + 2 , and D + 3 fractions of the beam were found to be a function of the gas throughput; at similar throughputs, the two gas feed systems produced similar extracted ion fractions. Operation at the proposed gas efficient point results in a small reduction (relative to the old high throughput mode) in the extracted D + fraction of the beam from 77% to 71%, with concomitant changes in the D + 2 fraction from 18% to 26%, and 6% to 3% for D + 3 . The injected power is unchanged, ∼2.2 MW at 95 kV

Tritium Issues in Next Step Devices

Energy Technology Data Exchange (ETDEWEB)

C.H. Skinner; G. Federici

2001-09-05

Tritium issues will play a central role in the performance and operation of next-step deuterium-tritium (DT) burning plasma tokamaks and the safety aspects associated with tritium will attract intense public scrutiny. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition. Erosion of plasma-facing components will scale up with the pulse length from being barely measurable on existing machines to centimeter scale. Magnetic Fusion Energy (MFE) devices with carbon plasma-facing components will accumulate tritium by co-deposition with the eroded carbon and this will strongly constrain plasma operations. We report on a novel laser-based method to remove co-deposited tritium from carbon plasma-facing components in tokamaks. A major fraction of the tritium trapped in a co-deposited layer during the deuterium-tritium (DT) campaign on the Tokamak Fusion Test Reactor (TFTR) was released by heating with a scanning laser beam. This technique offers the potential for tritium removal in a next-step DT device without the use of oxidation and the associated deconditioning of the plasma-facing surfaces and expense of processing large quantities of tritium oxide. The operational lifetime of alternative materials such as tungsten has significant uncertainties due to melt layer loss during disruptions. Production of dust and flakes will need careful monitoring and minimization, and control and accountancy of the tritium inventory will be critical issues. Many of the tritium issues in Inertial Fusion Energy (IFE) are similar to MFE, but some, for example those associated with the target factory, are unique to IFE. The plasma-edge region in a tokamak has greater complexity than the core due to lack of poloidal symmetry and nonlinear feedback between the plasma and wall. Sparse diagnostic coverage and low dedicated experimental run time has hampered the

Tritium Issues in Next Step Devices

International Nuclear Information System (INIS)

C.H. Skinner; G. Federici

2001-01-01

Tritium issues will play a central role in the performance and operation of next-step deuterium-tritium (DT) burning plasma tokamaks and the safety aspects associated with tritium will attract intense public scrutiny. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition. Erosion of plasma-facing components will scale up with the pulse length from being barely measurable on existing machines to centimeter scale. Magnetic Fusion Energy (MFE) devices with carbon plasma-facing components will accumulate tritium by co-deposition with the eroded carbon and this will strongly constrain plasma operations. We report on a novel laser-based method to remove co-deposited tritium from carbon plasma-facing components in tokamaks. A major fraction of the tritium trapped in a co-deposited layer during the deuterium-tritium (DT) campaign on the Tokamak Fusion Test Reactor (TFTR) was released by heating with a scanning laser beam. This technique offers the potential for tritium removal in a next-step DT device without the use of oxidation and the associated deconditioning of the plasma-facing surfaces and expense of processing large quantities of tritium oxide. The operational lifetime of alternative materials such as tungsten has significant uncertainties due to melt layer loss during disruptions. Production of dust and flakes will need careful monitoring and minimization, and control and accountancy of the tritium inventory will be critical issues. Many of the tritium issues in Inertial Fusion Energy (IFE) are similar to MFE, but some, for example those associated with the target factory, are unique to IFE. The plasma-edge region in a tokamak has greater complexity than the core due to lack of poloidal symmetry and nonlinear feedback between the plasma and wall. Sparse diagnostic coverage and low dedicated experimental run time has hampered the

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

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

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

Annual report, July 1, 1975--September 30, 1976

International Nuclear Information System (INIS)

1976-01-01

This annual report gives brief descriptions of each of the following projects: (1) PLT, (2) PDX, (3) TFTR, (4) driven tokamak plasmas, (5) FM-1, (6) ATC, (7) linear devices, (8) reactor studies, (9) theory, (10) library, (11) planning, (12) utilities, (13) administration, (14) engineering, (15) machine design and fabrication, (16) graduate education, and (17) statistics

Princeton Plasma Physics Laboratory annual report, October 1, 1983-September 30, 1984

Energy Technology Data Exchange (ETDEWEB)

Phillips, C.A. (ed.)

1984-01-01

Progress made during this reporting period is reported for each of the following areas: (1) principal parameters achieved in experimental devices, (2) TFTR, (3) PLT, (4) PBX, (5) S-1 Spheromak, (6) advanced concepts Torus-1, (7) x-ray laser studies, (8) theory, (9) tokamak modeling, (10) reactor studies, (11) spin-polarized fusion program, (12) tokamak fusion core experiment, and (13) engineering. (MOW)

Princeton Plasma Physics Laboratory annual report, October 1, 1983-September 30, 1984

International Nuclear Information System (INIS)

Phillips, C.A.

1984-01-01

Progress made during this reporting period is reported for each of the following areas: (1) principal parameters achieved in experimental devices, (2) TFTR, (3) PLT, (4) PBX, (5) S-1 Spheromak, (6) advanced concepts Torus-1, (7) x-ray laser studies, (8) theory, (9) tokamak modeling, (10) reactor studies, (11) spin-polarized fusion program, (12) tokamak fusion core experiment, and (13) engineering

Encoding technique for high data compaction in data bases of fusion devices

International Nuclear Information System (INIS)

Vega, J.; Cremy, C.; Sanchez, E.; Portas, A.; Dormido, S.

1996-01-01

At present, data requirements of hundreds of Mbytes/discharge are typical in devices such as JET, TFTR, DIII-D, etc., and these requirements continue to increase. With these rates, the amount of storage required to maintain discharge information is enormous. Compaction techniques are now essential to reduce storage. However, general compression techniques may distort signals, but this is undesirable for fusion diagnostics. We have developed a general technique for data compression which is described here. The technique, which is based on delta compression, does not require an examination of the data as in delayed methods. Delta values are compacted according to general encoding forms which satisfy a prefix code property and which are defined prior to data capture. Several prefix codes, which are bit oriented and which have variable code lengths, have been developed. These encoding methods are independent of the signal analog characteristics and enable one to store undistorted signals. The technique has been applied to databases of the TJ-I tokamak and the TJ-IU torsatron. Compaction rates of over 80% with negligible computational effort were achieved. Computer programs were written in ANSI C, thus ensuring portability and easy maintenance. We also present an interpretation, based on information theory, of the high compression rates achieved without signal distortion. copyright 1996 American Institute of Physics

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

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

Present status of fusion researches in USA, 4

International Nuclear Information System (INIS)

Yoshikawa, Shoichi; Okabayashi, Michio

1983-01-01

25 years have elapsed since nuclear fusion was published at the second Geneva conference in 1958. During this period, the Plasma Physics Laboratory of Princeton University has achieved the central role in the research on toroidal system nuclear fusion devices. Also the experiment of the large tokamak TFTR started from December, 1982, recorded the longest containment time of 200 ms as the initial data, and toroidal devices look to approach one step close to the scientific verification experiment (Q = 1) of reactors. In the PPPL, in order to perfect the basis required for the realization of nuclear fusion reactors, the experimental and theoretical developments have been carried out. Plasma containment experiment has been advanced successively from stellarater through internal conductor type to tokamak, and in plasma heating, ion cyclotron heating, fast neutral particle injection heating and low region hybrid heating were successfully carried out. As the experimental apparatuses, that for poloidal divertor experiment, Princeton large torus, tokamak fusion test reactor (TFTR) and S-1 spheromak are described. From the theories developed recently, bean type tokamak, heliac-stellarator and nuclear fusion reaction utilizing μ-mesons and nuclear spin are explained. (Kako, I.)

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

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

Princeton Plasma Physics Laboratory Annual Site Environmental Report for Calendar Year 1997

Energy Technology Data Exchange (ETDEWEB)

Finley, V.L. and Levine, J.D.

1999-01-10

The results of the 1997 environmental surveillance and monitoring program for the Princeton Plasma Physics Laboratory (PPPL) are presented and discussed. The purpose of this report is to provide the U.S. Department of Energy and the public with information on the level of radioactive and non-radioactive pollutants, if any, that are added to the environment as a result of PPPL's operations. During Calendar Year 1997, PPPL's Tokamak Fusion Test Reactor (TFTR) completed fifteen years of fusion experiments begun in 1982. Over the course of three and half years of deuterium-tritium (D-T) plasma experiments, PPPL set a world record of 10.7 million watts of controlled fusion power, more than 700 tritium shots pulsed into the reactor vessel generating more than 5.6 x 10²⁰ neutron and 1.6 gigajoules of fusion energy and researchers studied plasma science experimental data, which included "enhanced reverse shear techniques." As TFTR was completing its historic operations, PPPL participated with the Oak Ridge National Laboratory, Columbia University, and the University of Washington (Seattle) in a collaboration effort to design the National Spherical Torus Experiment (NSTX). This next device, NSTX, is located in the former TFTR Hot Cell on D site, and it is designed to be a smaller and more economical torus fusion reactor. Construction of this device began in late 1997, and first plasma in scheduled for early 1999. For 1997, the U.S. Department of Energy in its Laboratory Appraisal report rated the overall performance of Princeton Plasma Physics Laboratory as "excellent." The report cited the Laboratory's consistently excellent scientific and technological achievements and its successful management practices, which included high marks for environmental management, employee health and safety, human resources administration, science education, and communications. Groundwater investigations continued under a voluntary agreement with the New Jersey

A three-barrel repeating pneumatic pellet injector for plasma fueling of the Joint European Torus

International Nuclear Information System (INIS)

Combs, S.K.; Milora, S.L.; Baylor, L.R.; Foust, C.R.; Gethers, F.E.; Sparks, D.O.

1987-01-01

Pellet fueling, the injection of frozen hydrogen isotope pellets at high velocity, has been used to improve plasma performance in various tokamak experiments. In one recent experiment, the repeating pneumatic hydrogen pellet injector was used on the Tokamak Fusion Test Reactor (TFTR). This machine gun-like device, which was developed at the Oak Ridge National Laboratory (ORNL) with an objective of steady-state fueling applications, was characterized by a fixed pellet size and a maximum repetition rate of 4 to 6 Hz for several seconds. It was used to deliver deuterium pellets at speeds ranging from 1.0 to 1.5 km/s into TFTR plasma discharges. In the first experiments, injection of single, large (nominal 4-mm-diam) pellets provided high plasma densities in TFTR (1.8 x 10 14 cm -3 on axis). After a conversion to smaller (nominal 2.7-mm-diam) pellets, the pellet injector was operated in the repeating mode to gradually increase the plasma density, injecting up to five pellets on a single machine pulse. This resulted in central plasma densities approaching 4 x 10 14 cm -3 and n tau values of 1.4 x 10 14 cm -3 s. For plasma fueling applications on the Joint European Torus (JET), a pellet injector fashioned after the prototype repeating pneumatic design has been developed. The versatile injector features three repeating guns in a common vacuum enclosure; the guns provide pellets that are 2.7, 4.0, and 6.0 mm in diameter and can operate independently at repetition rates of 5, 2.5, and 1 Hz, respectively. The injector has been installed on JET. A description of the equipment is presented, emphasizing the differences from the original repeating device. Performance characteristics of the three pneumatic guns are also included

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

The First Decommissioning of a Fusion Reactor Fueled by Deuterium-Tritium

International Nuclear Information System (INIS)

Gentile, Charles A.; Perry, Erik; Rule, Keith; Williams, Michael; Parsells, Robert; Viola, Michael; Chrzanowski, James

2003-01-01

The Tokamak Fusion Test Reactor (TFTR) at the Plasma Physics Laboratory of Princeton University (PPPL) was the first fusion reactor fueled by a mixture of deuterium and tritium (D-T) to be decommissioned in the world. The decommissioning was performed over a period of three years and was completed safely, on schedule, and under budget. Provided is an overview of the project and detail of various factors which led to the success of the project. Discussion will cover management of the project, engineering planning before the project started and during the field work as it was being performed, training of workers in the field, the novel adaptation of tools from other industry, and the development of an innovative process for the use of diamond wire to segment the activated/contaminated vacuum vessel. The success of the TFTR decommissioning provides a viable model for the decommissioning of D-T burning fusion devices in the future

Large-scale cryopumping for controlled fusion

International Nuclear Information System (INIS)

Pittenger, L.C.

1977-01-01

Vacuum pumping by freezing out or otherwise immobilizing the pumped gas is an old concept. In several plasma physics experiments for controlled fusion research, cryopumping has been used to provide clean, ultrahigh vacua. Present day fusion research devices, which rely almost universally upon neutral beams for heating, are high gas throughput systems, the pumping of which is best accomplished by cryopumping in the high mass-flow, moderate-to-high vacuum regime. Cryopumping systems have been developed for neutral beam injection systems on several fusion experiments (HVTS, TFTR) and are being developed for the overall pumping of a large, high-throughput mirror containment experiment (MFTF). In operation, these large cryopumps will require periodic defrosting, some schemes for which are discussed, along with other operational considerations. The development of cryopumps for fusion reactors is begun with the TFTR and MFTF systems. Likely paths for necessary further development for power-producing reactors are also discussed

Large-scale cryopumping for controlled fusion

Energy Technology Data Exchange (ETDEWEB)

Pittenger, L.C.

1977-07-25

Vacuum pumping by freezing out or otherwise immobilizing the pumped gas is an old concept. In several plasma physics experiments for controlled fusion research, cryopumping has been used to provide clean, ultrahigh vacua. Present day fusion research devices, which rely almost universally upon neutral beams for heating, are high gas throughput systems, the pumping of which is best accomplished by cryopumping in the high mass-flow, moderate-to-high vacuum regime. Cryopumping systems have been developed for neutral beam injection systems on several fusion experiments (HVTS, TFTR) and are being developed for the overall pumping of a large, high-throughput mirror containment experiment (MFTF). In operation, these large cryopumps will require periodic defrosting, some schemes for which are discussed, along with other operational considerations. The development of cryopumps for fusion reactors is begun with the TFTR and MFTF systems. Likely paths for necessary further development for power-producing reactors are also discussed.

Effect of plasma physics on choices of first wall materials and structures for a thermonuclear reactor

International Nuclear Information System (INIS)

Meade, D.M.

1975-01-01

Impurity ions adversely affect the behavior of present-day tokamaks, and control of impurities is expected to be a key element in determining the feasibility of thermonuclear fusion reactors. The plasma-surface interactions for tokamaks and several techniques for controlling impurities are described. The plasma-surface problem of next generation devices PLT, PDX, DIII and TFTR is expected to be similar to those encountered in a reactor. For these devices calculations indicate that most of the particle energy efflux will be in the 1 keV region. Ironically this energy region has not yet been investigated thoroughly by the surface physicists

Possible incorporation of a dee-shaped vacuum vessel in Doublet III

International Nuclear Information System (INIS)

Davis, L.; Rawls, J.M.

1979-11-01

The design of Doublet III allows relative straightforward incorporation of any of a number of possible dee-configuration vacuum vessels that can serve as relevant size tests of reactor regime devices. Configurations simulating those of JET, ETF and INTOR with plasma areas larger than TFTR can be attained with significant physics parameter results. Such modifications to Doublet III could be incorporated into planned upgrade activites with operations beginning in 1984, early enough to influence the designs of ETF and INTOR and test the scaling laws, poloidal coil system, and impurity control systems proposed for these ignition devices

Remote manipulation techniques in the maintenance and repair of nuclear power plants

International Nuclear Information System (INIS)

Rininsland, H.; Boehme, G.

1986-01-01

Remote manipulation means the application of multi-purpose devices featuring high mobility and universal applicability. The paper describes such a remote manipulation system (manipulation vehicle MF1 and MF2, master-slave manipulator, TFTR maintenance manipulator) which can be used flexibly in the NNP during maintenance and repair and in incident and accident situations. Connecting elements and tools can be redesigned for remote manipulation to meet the specific application environments. (DG) [de

Device description and general physics requirements

International Nuclear Information System (INIS)

Neilson, G.H. Jr.

1992-01-01

To accomplish the dual goals set forth in the mission statement - determination of burning plasma physics and demonstration of fusion power production - requires a tokamak with special characteristics. A conceptual design for such a facility has been developed by the CIT/BPX Project team over a period of about 5 years. The process has drawn extensively upon the world tokamak physics data base as well as engineering experience gained on actual machines like Tokamak Fusion Test Reactor (TFTR) and the Alcators and on design studies for machines like the Long-Pulse Ignited Test experiment (LITE) and the Tokamak Fusion Core Experiment (TFCX). The Joint European Torus (JET) and Doublet III-D (DIII-D) experiments in particular have had a significant influence on the physics base. The resulting design incorporates features that have proven successful in tokamak experiments, combined with new features that are needed in the regime of high-power-density deuterium-tritium (D-T) fusion plasmas

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

Proceedings of the 1984 International Conference on plasma physics

International Nuclear Information System (INIS)

Tran, M.Q.; Verbeek, R.J.

1985-01-01

The 1984 ICPP, held in Lausanne, Switzerland, is the third biennial conference of the series ''International conferences on plasma physics''. A complete spectrum of current plasma physics from fusion devices to interstellar space was presented, even if most of the papers were of direct interest for fusion. The conference stressed the important role that ''basic plasma physics'' must play in fusion research. Recent theoretical and experimental developments in tokamaks, stellarators, mirrors, reversed field pinches, and other fusion devices were reported. The successful operation of two newly-built large tokamak devices, JET and TFTR, holds the promise that a host of new results of decisive importance for fusion research will become available in the next few years. This is the first part of the conference

Application of optical fluctuation diagnostics to transport studies in high temperature tokamak plasmas

International Nuclear Information System (INIS)

Fonck, R.J.

1991-01-01

First data was obtained from the TFTR BES system in May 1990, after the prototype 4-channel system with photomultiplier detectors from PBX-M was transferred to the TFTR Hot Cell. The system was expanded to 16 spatial channels during the Summer 1990 run period, and data of interest was obtained near the end of the run in October 1990. The most attention has been given to a radial profile data set obtained for an L-mode plasma condition in TFTR, where we characterized the nature of the long-wavelength turbulence seen by the BES diagnostic. We summarize here highlights of the initial run with BES or TFTR. We also give a progress report on support activities at UW

The tokamak fusion test reactor tritium systems test contractor operational readiness review

International Nuclear Information System (INIS)

Gentile, C.A.; Levine, J.; Norris, M.; Rehill, F.; Such, C.

1993-01-01

In preparation for D-T operations at TFTR, the TFTR project has successfully completed the C-ORR process which has led to the introduction of 200 curies of tritium to the site. Preparations for the C-ORR began approximately 2 years ago. During July 1992 a one-week Site Assistance Review was conducted by the C-ORR , and C-ORR Team consisting of 12 persons, all of whom were outside experts, many of whom were from other facilities within the DOE complex. During the July 1992 Site Assistance Review 201 findings were documented which fell into one of three categories. All of the 109 category one findings which were generated were required to be resolved prior to the introduction of tritium to the TFTR site. On April 5, 1993, the TFTR Tritium System Test C-ORR commenced. The results of the C-ORR as documented in the final report by the C-ORR was that category 1 findings were resolved, and it was the recommendation of the C-ORR Team to the PPPL ES ampersand H Board that TFTR initiate the Tritium Systems Test. DOE (Chicago Operations, Princeton Area Office) concurred with the C-ORR final report and on April 29, 1993, at 12:15 pm tritium was introduced to the TFTR site

Review of D-T Experiments Relevant to Burning Plasma Issues

International Nuclear Information System (INIS)

Hawryluk, R.J.

2001-01-01

Progress in the performance of tokamak devices has enabled not only the production of significant bursts of fusion energy from deuterium-tritium (D-T) 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. The TFTR and JET, in conjunction with the worldwide fusion effort, have studied a broad range of topics including magnetohydrodynamic stability, transport, wave-particle interactions, the confinement of energetic particles, and plasma boundary interactions. The D-T experiments differ in three principal ways from previous experiments: isotope effects associated with the use of deuterium-tritium fuel, the presence of fusion-generated alpha particles, and technology issues associated with tritium handling and increased activation. The effect of deuterium-tritium fuel and the presence of alpha particles is reviewed and placed in the perspective of the much large r worldwide database using deuterium fuel and theoretical understanding. Both devices have contributed substantially to addressing the scientific and technical issues associated with burning plasmas. However, future burning plasma experiments will operate with larger ratios of alpha heating power to auxiliary power and will be able to access additional alpha-particle physics issues. The scientific opportunities for extending our understanding of burning plasmas beyond that provided by current experiments is described

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

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

Fusion Concept Exploration Experiments at PPPL

International Nuclear Information System (INIS)

Stewart Zweben; Samuel Cohen; Hantao Ji; Robert Kaita; Richard Majeski; Masaaki Yamada

1999-01-01

Small ''concept exploration'' experiments have for many years been an important part of the fusion research program at the Princeton Plasma Physics Laboratory (PPPL). this paper describes some of the present and planned fusion concept exploration experiments at PPPL. These experiments are a University-scale research level, in contrast with the larger fusion devices at PPPL such as the National Spherical Torus Experiment (NSTX) and the Tokamak Fusion Test Reactor (TFTR), which are at ''proof-of-principle'' and ''proof-of-performance'' levels, respectively

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''

Pellet injectors for the tokamak fusion test reactor

International Nuclear Information System (INIS)

Combs, S.K.

1986-01-01

The repeating pneumatic injector is a device from the ORNL development program. A new eight-shot deuterium pellet injector has been designed and constructed specifically for the TFTR application and is scheduled to replace the repeating injector this year. The new device combines a cryogenic extruder and a cold wheel rotary mechanism to form and chamber eight pellets in a batch operation; the eight pellets can then be delivered in any time sequence. Another unique feature of the device is the variable pellet size with three pellets each of 3.0 and 3.5 mm diam and two each of 4.0 mm diam. The experience and technology that have been developed on previous injectors at ORNL have been utilized in the design of this latest pellet injection system

Radiation environment of fusion experimental reactor

International Nuclear Information System (INIS)

Mori, Seiji; Seki, Yasushi

1988-01-01

Next step device (experimental reactor), which is planned to succeed the large plasma experimental devices such as JT-60, JET and TFTR, generates radiation (neutron + gamma ray) during its operation. Radiation (neutronic) properties of the material are basis for the study on neutron utilization (energy recovery and tritium breeding), material selection (irradiation damage and lifetime evaluation) and radiation safety (personnel exposure and radiation waste). It is necessary, therefore, to predict radiation behaviour in the reactor correctly for the engineering design of the reactor. This report describes the outline of the radiation environment of the reactor based on the information obtained by the neutronic and shielding design calculation of the fusion experimental reactor (FER). (author)

Improvement in Plasma Performance with Lithium Coatings in NSTX

International Nuclear Information System (INIS)

Kaita, R.

2009-01-01

Lithium as a plasma-facing material has attractive features, including a reduction in the recycling of hydrogenic species and the potential for withstanding high heat and neutron fluxes in fusion reactors. Dramatic effects on plasma performance with lithium-coated plasma-facing components (PFC's) have been demonstrated on many fusion devices, including TFTR, T-11M, and FT-U. Using a liquid-lithium-filled tray as a limiter, the CDX-U device achieved very significant enhancement in the confinement time of ohmically heated plasmas. The recent NSTX experiments reported here have demonstrated, for the first time, significant and recurring benefits of lithium PFC coatings on divertor plasma performance in both L- and H- mode regimes heated by neutral beams.

In-vessel maintenance remote manipulator system

International Nuclear Information System (INIS)

Jimenez, E.

1978-01-01

The radiation environment within the Tokamak Fusion Test Reactor (TFTR) vacuum vessel necessitates the development of a Remote Manipulator System (RMS) to perform required periodic inspection and maintenance tasks. The RMS must be able to perform dexterous operations and handle loads that exceed human capabilities. The limited size of the access ports on the TFTR vacuum vessel and the performance profile, defined by the various handling requirements, present unique design constraints. The design approach and formulation of a RMS configuration which satisfies TFTR requirements is presented herein

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

TPX assembly plan

International Nuclear Information System (INIS)

Knutson, D.

1993-01-01

The TPX machine will be assembled in the TFTR Test Cell at the Plasma Physics Laboratory, utilizing the existing TFTR machine foundation. Preparation of the area for assembly will begin after completion of the decontamination and decommissioning phase on TFTR and certification that the radiation levels remaining, if any, are consistent with the types of operations planned. Assembly operations begin with the arrival of the first components, and conclude, approximately 24 months later, with the successful completion of the integrated systems tests and the achievement of a first plasma

Present status of plasma-wall interactions research and materials development activities in the US

International Nuclear Information System (INIS)

Hirooka, Y.; Conn, R.W.

1989-08-01

It is well known in the fusion engineering community that the plasma confinement performance in magnetic fusion devices is strongly affected by edge-plasma interactions with surface components. These plasma-material interactions (PMI) include fuel particle recycling and impurity generation both during normal and off-normal operation. To understand and then to control PMI effects, considerable effort has been made, particularly over the last decade in US, supported by Department of Energy, Division of Development and Technology. Also, because plasma-facing components are generally expected to receive significant amount of heat due to plasma bombardment and run-away electrons, materials must tolerate high-heat fluxes (HHF). The HHF-component research has been conducted in parallel with PMI research. One strong motivation for these research activities is that DT-burning experiments are currently planned in the Tokamak Test Fusion Reactor (TFTR) in early 1990s. Several different but mutually complementary approaches have been taken in the PMI+HHF research. The first approach is to conduct PMI experiments using toroidal fusion devices such as TFTR. The second one is to simulate elemental processes involved in PMI using ion beams and electron beams, etc. The last one but not least is to use non-tokamak plasma facilities. Along with these laboratory activities, new materials have been developed and evaluated from the PMI+HHF point of view. In this paper, several major PMI+HHF research facilities in US and their activities are briefly reviewed. 21 refs., 10 figs., 2 tabs

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

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

Tritium Removal from Carbon Plasma Facing Components

International Nuclear Information System (INIS)

Skinner, C.H.; Coad, J.P.; Federici, G.

2003-01-01

Tritium removal is a major unsolved development task for next-step devices with carbon plasma-facing components. The 2-3 order of magnitude increase in duty cycle and associated tritium accumulation rate in a next-step tokamak will place unprecedented demands on tritium removal technology. The associated technical risk can be mitigated only if suitable removal techniques are demonstrated on tokamaks before the construction of a next-step device. This article reviews the history of codeposition, the tritium experience of TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) and the tritium removal rate required to support ITER's planned operational schedule. The merits and shortcomings of various tritium removal techniques are discussed with particular emphasis on oxidation and laser surface heating

The present role of superconductivity in fusion

International Nuclear Information System (INIS)

Shimamoto, S.

1986-01-01

After completion of large fusion devices in the world, such as JT-60, JET and TFTR, high temperature plasma is proceeding to critical condition for fusion. The devices up to now use mainly conventional magnet. However, for the next generation machine which demonstrates fusion reaction, deuterium-tritium burning, superconducting magnet system is indispensable from view point of both net energy extraction and capacity limitation of power supply. In order to realize such a large and complicated system, a lot of development works is being carried out. This paper describes required parameters of superconducting magnet and helium refrigerator, the state of plasma condition and superconducting magnet. It is shown that the present technology of superconducting magnet is not so far from realization of fusion experimental reactor

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

Protection device for a thermonuclear device

International Nuclear Information System (INIS)

Kawashima, Shuichi.

1986-01-01

Purpose: To exactly detect the void coefficients of coolants even under high magnetic fields thereby detect the overheat of a thermonuclear device at an early stage. Constitution: The protecting device of this invention comprises a laser beam generation device, a laser beam detection device and an accident detection device. The laser generation device always generates laser beams, which are permeated through coolants and detected by the laser beam detection device, the optical amount of which is transmitted to the accident detection device. The accident detection device judges the excess or insufficiency of the detected optical amount with respect to the optical amount of the laser beams under the stationary state as a reference and issues an accident signal. Since only the optical cables that do not undergo the effect of the magnetic fields are exposed to high magnetic fields in the protection device of this invention, a high reliability can be maintained. (Kamimura, M.)

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

Device-Centric Monitoring for Mobile Device Management

Directory of Open Access Journals (Sweden)

Luke Chircop

2016-03-01

Full Text Available The ubiquity of computing devices has led to an increased need to ensure not only that the applications deployed on them are correct with respect to their specifications, but also that the devices are used in an appropriate manner, especially in situations where the device is provided by a party other than the actual user. Much work which has been done on runtime verification for mobile devices and operating systems is mostly application-centric, resulting in global, device-centric properties (e.g. the user may not send more than 100 messages per day across all applications being difficult or impossible to verify. In this paper we present a device-centric approach to runtime verify the device behaviour against a device policy with the different applications acting as independent components contributing to the overall behaviour of the device. We also present an implementation for Android devices, and evaluate it on a number of device-centric policies, reporting the empirical results obtained.

Determination of atomic data pertinent to the Magnetic Fusion Program: Technical progress report, 15 May 1986-30 September 1987

International Nuclear Information System (INIS)

Wiese, W.L.

1987-01-01

Dielectronic recombination and excitation rates, electron-impact excitation and ionization cross sections, and wavelengths and energy levels of prominent spectral lines are experimentally and theoretically determined. Wavelengths for both electric and magnetic dipole transitions and atomic energy level data are also critically evaluated, compiled, and tabulated. Theoretical methods use both relativistic and nonrelativistic formulations. The work concentrated on ions of materials commonly used in current fusion devices, such as titanium, iron, and nickel, as well as heavier elements expected to be introduced into next-generation fusion devices for diagnostic purposes, such as krypton and xenon. The range of ions is extended to include very highly charged species in anticipation of needs in very high-temperature fusion devices such as TFTR and its successors. Work described also represents collaboration with major fusion laboratories such as Oak Ridge National Laboratory, Princeton Plasma Physics Laboratory, and GA Technologies

Continued development of modeling tools and theory for RF heating

International Nuclear Information System (INIS)

1998-01-01

Mission Research Corporation (MRC) is pleased to present the Department of Energy (DOE) with its renewal proposal to the Continued Development of Modeling Tools and Theory for RF Heating program. The objective of the program is to continue and extend the earlier work done by the proposed principal investigator in the field of modeling (Radio Frequency) RF heating experiments in the large tokamak fusion experiments, particularly the Tokamak Fusion Test Reactor (TFTR) device located at Princeton Plasma Physics Laboratory (PPPL). An integral part of this work is the investigation and, in some cases, resolution of theoretical issues which pertain to accurate modeling. MRC is nearing the successful completion of the specified tasks of the Continued Development of Modeling Tools and Theory for RF Heating project. The following tasks are either completed or nearing completion. (1) Anisotropic temperature and rotation upgrades; (2) Modeling for relativistic ECRH; (3) Further documentation of SHOOT and SPRUCE. As a result of the progress achieved under this project, MRC has been urged to continue this effort. Specifically, during the performance of this project two topics were identified by PPPL personnel as new applications of the existing RF modeling tools. These two topics concern (a) future fast-wave current drive experiments on the large tokamaks including TFTR and (c) the interpretation of existing and future RF probe data from TFTR. To address each of these topics requires some modification or enhancement of the existing modeling tools, and the first topic requires resolution of certain theoretical issues to produce self-consistent results. This work falls within the scope of the original project and is more suited to the project's renewal than to the initiation of a new project

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

Princeton Plasma Physics Laboratory (PPPL) annual site environmental report for calendar year 1994

International Nuclear Information System (INIS)

Finley, V.L.; Wieczorek, M.A.

1996-02-01

This report gives the results of the environmental activities and monitoring programs at the Princeton Plasma Physics Laboratory (PPPL) for CY94. The report is prepared to provide the US Department of Energy (DOE) and the public with information on the level of radioactive and nonradioactive pollutants, if any, added to the environment as a result of PPPL operations, as well as environmental initiatives, assessments, and programs that were undertaken in 1994. The objective of the Annual Site Environmental Report is to document evidence that PPPL's environmental protection programs adequately protect the environment and the public health. The Princeton Plasma Physics Laboratory has engaged in fusion energy research since 195 1. The long-range goal of the US Magnetic Fusion Energy Research Program is to develop and demonstrate the practical application of fusion power as an alternate energy source. In 1994, PPPL had one of its two large tokamak devices in operation-the Tokamak Fusion Test Reactor (TFTR). The Princeton Beta Experiment-Modification or PBX-M completed its modifications and upgrades and resumed operation in November 1991 and operated periodically during 1992 and 1993; it did not operate in 1994 for funding reasons. In December 1993, TFTR began conducting the deuterium-tritium (D-T) experiments and set new records by producing over ten at sign on watts of energy in 1994. The engineering design phase of the Tokamak Physics Experiment (T?X), which replaced the cancelled Burning Plasma Experiment in 1992 as PPPL's next machine, began in 1993 with the planned start up set for the year 2001. In December 1994, the Environmental Assessment (EA) for the TFTR Shutdown and Removal (S ampersand R) and TPX was submitted to the regulatory agencies, and a finding of no significant impact (FONSI) was issued by DOE for these projects

The operation of the Tokamak Fusion Test Reactor Tritium Facility

International Nuclear Information System (INIS)

Gentile, C.A.; LaMarche, P.H.

1995-01-01

The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During '' time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems

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

Visual tritium imaging of In-Vessel surfaces

International Nuclear Information System (INIS)

Gentile, C. A.; Zweben, S. J.; Skinner, C. H.; Young, K. M.; Langish, S. W.; Nishi, M. F.; Shu, W. M.; Parker, J.; Isobe, K.

2000-01-01

A imaging detector has been developed for the purpose of providing a non-destructive, real time method of determining tritium concentrations on the surface of internal TFTR vacuum vessel components. The detector employs a green phosphor screen (P31, zinc sulfide: copper) with a wave length peak of 530 nm, a charge-coupled device (CCD) camera linked to a computer, and a detection chamber for inserting components recovered from the vacuum vessel. This detector is capable of determining tritium concentrations on the surfaces. The detector provides a method of imaging tritium deposition on the surfaces in a fairly rapid fashion

Visual tritium imaging of in-vessel surfaces

International Nuclear Information System (INIS)

Gentile, C.A.; Zweben, S.J.; Skinner, C.H.; Young, K.M.; Langish, S.W.; Nishi, M.F.; Shu, W.M.; Parker, J.; Isobe, K.

2000-01-01

An imaging detector has been developed for the purpose of providing a non-destructive, real time method of determining tritium concentrations on the surface of internal TFTR vacuum vessel components. The detector employs a green phosphor screen (P31, zinc sulfide: copper) with a wave length peak of 530 nm, a charge-coupled device (CCD) camera linked to a computer, and a detection chamber for inserting components recovered from the vacuum vessel. This detector is capable of determining tritium concentrations on the surfaces. The detector provides a method of imaging tritium deposition on the surfaces in a fairly rapid fashion

Power source device for thermonuclear device

International Nuclear Information System (INIS)

Ozaki, Akira.

1992-01-01

The present invention provides a small sized and economical power source device for a thermonuclear device. That is, the device comprises a conversion device having a rated power determined by a power required during a plasma current excitation period and a conversion device having a rated power determined by a power required during a plasma current maintaining period, connected in series to each other. Then, for the former conversion device, power is supplied from an electric power generator and, for the latter, power is supplied from a power system. With such a constitution, during the plasma electric current maintaining period for substantially continuous operation, it is possible to conduct bypassing paired operation for the former conversion device while the electric power generator is put under no load. Further, since a short period rated power may be suffice for the former conversion device and the electric power generator having the great rated power required for the plasma electric current excitation period, they can be reduced in the size and made economical. On the other hand, since the power required for the plasma current maintaining period is relatively small, the capacity of the continuous rated conversion device may be small, and the power can be received from the power system. (I.S.)

Software problems in magnetic fusion research

International Nuclear Information System (INIS)

Gruber, R.

1982-01-01

The main world effort in magnetic fusion research involves studying the plasma in a Tokamak device. Four large Tokamaks are under construction (TFTR in USA, JET in Europe, T15 in USSR and JT60 in Japan). To understand the physical phenomena that occur in these costly devices, it is generally necessary to carry out extensive numerical calculations. These computer simulations make use of sophisticated numerical methods and demand high power computers. As a consequence they represent a substantial investment. To reduce software costs, the computer codes are more and more often exhanged among scientists. Standardization (STANDARD FORTRAN, OLYMPUS system) and good documentation (CPC program library) are proposed to make codes exportable. Centralized computing centers would also help in the exchange of codes and ease communication between the staff at different laboratories. (orig.)

Energetic Particle Physics In Fusion Research In Preparation For Burning Plasma Experiments

Energy Technology Data Exchange (ETDEWEB)

Gorelenkov, Nikolai N [PPPL

2013-06-01

The area of energetic particle (EP) physics of fusion research has been actively and extensively researched in recent decades. The progress achieved in advancing and understanding EP physics has been substantial since the last comprehensive review on this topic by W.W. Heidbrink and G.J. Sadler [1]. That review coincided with the start of deuterium-tritium (DT) experiments on Tokamak Fusion Test reactor (TFTR) and full scale fusion alphas physics studies. Fusion research in recent years has been influenced by EP physics in many ways including the limitations imposed by the "sea" of Alfven eigenmodes (AE) in particular by the toroidicityinduced AEs (TAE) modes and reversed shear Alfven (RSAE). In present paper we attempt a broad review of EP physics progress in tokamaks and spherical tori since the first DT experiments on TFTR and JET (Joint European Torus) including helical/stellarator devices. Introductory discussions on basic ingredients of EP physics, i.e. particle orbits in STs, fundamental diagnostic techniques of EPs and instabilities, wave particle resonances and others are given to help understanding the advanced topics of EP physics. At the end we cover important and interesting physics issues toward the burning plasma experiments such as ITER (International Thermonuclear Experimental Reactor).

Device-Centric Monitoring for Mobile Device Management

OpenAIRE

Chircop, Luke; Colombo, Christian; Pace, Gordon J.

2016-01-01

The ubiquity of computing devices has led to an increased need to ensure not only that the applications deployed on them are correct with respect to their specifications, but also that the devices are used in an appropriate manner, especially in situations where the device is provided by a party other than the actual user. Much work which has been done on runtime verification for mobile devices and operating systems is mostly application-centric, resulting in global, device-centri...

NSTX Electrical Power Systems

International Nuclear Information System (INIS)

A. Ilic; E. Baker; R. Hatcher; S. Ramakrishnan; et al

1999-01-01

The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physic Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. The design of the NSTX electrical power system was tailored to suit the available infrastructure and electrical equipment on site. Components were analyzed to verify their suitability for use in NSTX. The total number of circuits and the location of the NSTX device drove the major changes in the Power system hardware. The NSTX has eleven (11) circuits to be fed as compared to the basic three power loops for TFTR. This required changes in cabling to insure that each cable tray system has the positive and negative leg of cables in the same tray. Also additional power cabling had to be installed to the new location. The hardware had to b e modified to address the need for eleven power loops. Power converters had to be reconnected and controlled in anti-parallel mode for the Ohmic heating and two of the Poloidal Field circuits. The circuit for the Coaxial Helicity Injection (CHI) System had to be carefully developed to meet this special application. Additional Protection devices were designed and installed for the magnet coils and the CHI. The thrust was to making the changes in the most cost-effective manner without compromising technical requirements. This paper describes the changes and addition to the Electrical Power System components for the NSTX magnet systems

Suggestions for an updated fusion power program

International Nuclear Information System (INIS)

Clarke, J.F.

1976-02-01

This document contains suggestions for a revised CTR Program strategy which should allow us to achieve equivalent goals while operating within the above constraints. The revised program is designed around three major facilities. The first is an upgrading of the present TFTR facility which will provide a demonstration of the generation of tens of megawatts electric equivalent originally envisioned for the 1985 EPR. The second device is the TTAP which will allow the integration and optimization of the plasma physics results obtained from the next generation of plasma physics experiments. The improvement in tokamak reactor operation resulting from this optimization of fusion plasma performance will enable an EPR to be designed which will produce several hundred megawatts of electric power by 1990. This will move the fusion program much closer to its goal of commercial fusion power by the turn of the century. In addition to this function the TTAP will serve as a prototype of the 1990 EPR system, thus making more certain the successful operation of this device. The third element of this revised program is an intense radiation damage facility which will provide the radiation damage information necessary for the EPR and subsequent fusion reactor facilities. The sum total of experience gained from reacting plasma experiments on TFTR, reactor grade plasma optimization and technological prototyping on TTAP, and end of life radiation damage results from the intense neutron facility will solve all of the presently foreseen problems associated with a tokamak fusion power reactor except those associated with the external nuclear systems. These external system problems such as tritium breeding and optimal power recovery can be developed in parallel on the 1990 EPR

IAEA INTOR workshop report, groups 2, 5, 7, 9, 10 and 15

International Nuclear Information System (INIS)

1980-02-01

In order to prove scientific feasibility of magnetic confinement fusion, large fusion devices are under construction in several countries (JT-60 in Japan, T-15 in U.S.S.R., TFTR in U.S.A. and JET in EC). International Tokamak Reactor (INTOR) Workshop was organized by the International Atomic Energy Agency (IAEA) to identify roles, objectives and characteristics of the next generation fusion device. This report is a compilation of the home task reports of six groups on INTOR engineering aspects by Japan Atomic Energy Research Institute for workshop sessions 2 and 3 held in 1979. Tasks of the respective groups are group 2: first wall/blanket/shield, group 5: magnetics, group 7: systems integration and structure, group 9: assembly and remote maintenance, group 10: radiation shielding and personnel access, group 15: safety and environment. (author)

High beta and second stability region transport and stability analysis. Final report

Energy Technology Data Exchange (ETDEWEB)

Hughes, M.H.; Phillips, M.W.

1996-01-01

This report describes MHD equilibrium and stability studies carried out at Northrop Grumman`s Advanced Technology and Development Center during the period March 1 to December 31, 1995. Significant progress is reported in both ideal and resistive MHD modeling of TFTR plasmas. Specifically, attention is concentrated on analysis of Advanced Tokamak experiments at TFTR involving plasmas in which the q-profiles were non-monotonic.

A survey of reflectometry techniques with applications to TFTR

International Nuclear Information System (INIS)

Collazo, I.; Stacey, W.M.; Wilgen, J.; Hanson, G.; Bigelow, T.; Thomas, C.E.; Bretz, N.

1993-12-01

This report presents a review of reflectometry with particular attention to eXtraordinary mode (X-mode) reflectometry using the novel technique of dual frequency differential phase. The advantage of using an X-mode wave is that it can probe the edge of the plasma with much higher resolution and using a much smaller frequency range than with the Ordinary mode (O-Mode). The general problem with previous full phase reflectometry techniques is that of keeping track of the phase (on the order of 1000 fringes) as the frequency is swept over the band. The dual frequency phase difference technique has the advantage that since it is keeping track of the phase difference of two frequencies with a constant frequency separation, the fringe counting is on the order of only 3 to 5 fringes. This fringe count, combined with the high resolution of the X-mode wave and the small plasma access requirements of reflectometry, make X-mode reflectometry a very attractive diagnostic for today's experiments and future fusion devices

Ontology-Based Device Descriptions and Device Repository for Building Automation Devices

Directory of Open Access Journals (Sweden)

Dibowski Henrik

2011-01-01

Full Text Available Device descriptions play an important role in the design and commissioning of modern building automation systems and help reducing the design time and costs. However, all established device descriptions are specialized for certain purposes and suffer from several weaknesses. This hinders a further design automation, which is strongly needed for the more and more complex building automation systems. To overcome these problems, this paper presents novel Ontology-based Device Descriptions (ODDs along with a layered ontology architecture, a specific ontology view approach with virtual properties, a generic access interface, a triple store-based database backend, and a generic search mask GUI with underlying query generation algorithm. It enables a formal, unified, and extensible specification of building automation devices, ensures their comparability, and facilitates a computer-enabled retrieval, selection, and interoperability evaluation, which is essential for an automated design. The scalability of the approach to several ten thousand devices is demonstrated.

FY93 Princeton Plasma Physics Laboratory. Annual report, October 1, 1992--September 30, 1993

Energy Technology Data Exchange (ETDEWEB)

1995-02-01

This is the annual report from the Princeton Plasma Physics Laboratory for the period October 1, 1992 to September 30, 1993. The report describes work done on TFTR during the year, as well as preparatory to beginning of D-T operations. Design work is ongoing on the Tokamak Physics Experiment (TPX) which is to test very long pulse operations of tokamak type devices. PBX has come back on line with additional ion-Bernstein power and lower-hybrid current drive. The theoretical program is also described, as well as other small scale programs, and the growing effort in collaboration on international design projects on ITER and future collaborations at a larger scale.

FY93 Princeton Plasma Physics Laboratory. Annual report, October 1, 1992--September 30, 1993

International Nuclear Information System (INIS)

1995-01-01

This is the annual report from the Princeton Plasma Physics Laboratory for the period October 1, 1992 to September 30, 1993. The report describes work done on TFTR during the year, as well as preparatory to beginning of D-T operations. Design work is ongoing on the Tokamak Physics Experiment (TPX) which is to test very long pulse operations of tokamak type devices. PBX has come back on line with additional ion-Bernstein power and lower-hybrid current drive. The theoretical program is also described, as well as other small scale programs, and the growing effort in collaboration on international design projects on ITER and future collaborations at a larger scale

Design studies of supplemental-heating systems. Quarterly report No. 2

International Nuclear Information System (INIS)

Arnold, K.W.; Birnbaum, D.N.; Cole, A.J.; Freije, S.A.; Romesser, T.E.; Samec, T.K.; Thomson, J.J.

1981-01-01

The first objective is to summarize and evaluate the data base for Tokamak heating systems, in order to understand thoroughly the technology, cost and risk implications associated with the selection of any particular heating technique. A preliminary selection of optimized RF and neutral beam systems will be made for a near-term device (TFTR), based on existing data. This will provide a technical and cost base for the second phase where concept and design selection will be performed for an advanced reactor system. In addition, the analysis procedure will be carefully documented, together with a program plan for the development of each of the systems and a detailed cost estimate

Multi-Device to Multi-Device (MD2MD Content-Centric Networking Based on Multi-RAT Device

Directory of Open Access Journals (Sweden)

Cheolhoon Kim

2017-11-01

Full Text Available This paper proposes a method whereby a device can transmit and receive information using a beacon, and also describes application scenarios for the proposed method. In a multi-device to multi-device (MD2MD content-centric networking (CCN environment, the main issue involves searching for and connecting to nearby devices. However, if a device can’t find another device that satisfies its requirements, the connection is delayed due to the repetition of processes. It is possible to rapidly connect to a device without repetition through the selection of the optimal device using the proposed method. Consequently, the proposed method and scenarios are advantageous in that they enable efficient content identification and delivery in a content-centric Internet of Things (IoT environment, in which multiple mobile devices coexist.

Development of Long-Pulse Heating and Current Drive Actuators and Operational Techniques Compatible with a High-Z Divertor and First Wall

Energy Technology Data Exchange (ETDEWEB)

Wang, Guiding [Univ. of California, Los Angeles, CA (United States)

2017-11-02

Accurate measurement of the edge electron density profile is essential to optimizing antenna coupling and assessment of impurity contamination in studying long-pulse plasma heating and current drive in fusion devices. Measurement of the edge density profile has been demonstrated on the US fusion devices such as C-Mod, DIII-D, and TFTR amongst many devices, and has been used for RF loading and impurity modeling calculations for many years. University of Science and Technology of China (USTC) has recently installed a density profile reflectometer system on the EAST fusion device at the Institute of Plasma Physics, Chinese Academy of Sciences in China based on the University of California Los Angeles (UCLA)-designed reflectometer system on the DIII-D fusion device at General Atomics Company in San Diego, California. UCLA has been working with USTC to optimize the existing microwave antenna, waveguide system, microwave electronics, and data analysis to produce reliable edge density profiles. During the past budget year, progress has been made in all three major areas: effort to achieve reliable system operations under various EAST operational conditions, effort to optimize system performance, and effort to provide quality density profiles into EAST’s database routinely.

Graphene device and method of using graphene device

Science.gov (United States)

Bouchiat, Vincent; Girit, Caglar; Kessler, Brian; Zettl, Alexander K.

2015-08-11

An embodiment of a graphene device includes a layered structure, first and second electrodes, and a dopant island. The layered structure includes a conductive layer, an insulating layer, and a graphene layer. The electrodes are coupled to the graphene layer. The dopant island is coupled to an exposed surface of the graphene layer between the electrodes. An embodiment of a method of using a graphene device includes providing the graphene device. A voltage is applied to the conductive layer of the graphene device. Another embodiment of a method of using a graphene device includes providing the graphene device without the dopant island. A dopant island is placed on an exposed surface of the graphene layer between the electrodes. A voltage is applied to the conductive layer of the graphene device. A response of the dopant island to the voltage is observed.

Humanitarian Use Devices/Humanitarian Device Exemptions in cardiovascular medicine.

Science.gov (United States)

Kaplan, Aaron V; Harvey, Elisa D; Kuntz, Richard E; Shiran, Hadas; Robb, John F; Fitzgerald, Peter

2005-11-01

The Second Dartmouth Device Development Symposium held in October 2004 brought together leaders from the medical device community, including clinical investigators, senior representatives from the US Food and Drug Administration, large and small device manufacturers, and representatives from the financial community to examine difficult issues confronting device development. The role of the Humanitarian Use Device/Humanitarian Device Exemption (HUD/HDE) pathway in the development of new cardiovascular devices was discussed in this forum. The HUD/HDE pathway was created by Congress to facilitate the availability of medical devices for "orphan" indications, ie, those affecting HDEs have been granted (23 devices, 6 diagnostic tests). As the costs to gain regulatory approval for commonly used devices increase, companies often seek alternative ways to gain market access, including the HUD/HDE pathway. For a given device, there may be multiple legitimate and distinct indications, including indications that meet the HUD criteria. Companies must choose how and when to pursue each of these indications. The consensus of symposium participants was for the HUD/HDE pathway to be reserved for true orphan indications and not be viewed strategically as part of the clinical development plan to access a large market.

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

DeviceNet-based device-level control in SSRF

CERN Document Server

Leng Yong Bin; Lu Cheng Meng; Miao Hai Feng; Liu Song Qiang; Shen Guo Bao

2002-01-01

The control system of Shanghai Synchrotron Radiation Facility is an EPICS-based distributed system. One of the key techniques to construct the system is the device-level control. The author describes the design and implementation of the DeviceNet-based device controller. A prototype of the device controller was tested in the experiments of magnet power supply and the result showed a precision of 3 x 10 sup - sup 5

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

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

The scientific prototype, the only reasonable next step for the American MFE program; or why FESAC will never solicit my advice again

Science.gov (United States)

Manheimer, Wallace

2014-10-01

The scientific prototype is a tokamak which builds on what has been accomplished in TFTR, JET and JT-60. Instead of attempting to advance the plasma parameters, or investigate a new confinement configuration, it takes the tokamak plasma parameters already achieved (or actually nearly already achieved), Q about 1 and run it at steady state or high duty cycle in a DT plasma. It is very much a nuclear device requiring all of the safeguards of any nuclear device. It is an important step forward for either pure fusion or fusion breeding, and it is difficult to see how fusion can advance very far with out the knowledge the scientific prototype would provide. The poster will be divided into two parts. The first part examines options other than the scientific prototype and shows why they should be rejected. The second part explains the scientific prototype in somewhat more detail.

QoE-Aware Device-to-Device Multimedia Communications

Directory of Open Access Journals (Sweden)

Liang ZHOU

2015-08-01

Full Text Available Multimedia services over mobile device-to-device (D2D networks has recently received considerable attention. In this scenario, each device is equipped with a cellular communication interface, as well as a D2D interface over a shared medium. In this work, we study the performance properties of the mobile D2D communications in the framework of user satisfaction, and develop a fully distributed QoE-aware multimedia communication scheme (QAMCS. Specifically, we translate the opportunistic multimedia communications issue into a stochastic optimization problem, which opens up a new degree of performance to exploit. Moreover, QAMCS is designed for a heterogeneous and dynamic environment, in which user demand, device mobility, and transmission fashion may vary across different devices and applications. Importantly, QAMCS is able to maximize the user satisfaction and only needs each device to implement its own scheme individually in the absence of a central controller.

Virtual MIMO Beamforming and Device Pairing Enabled by Device-to-Device Communications for Multidevice Networks

Directory of Open Access Journals (Sweden)

Yeonjin Jeong

2017-01-01

Full Text Available We consider a multidevice network with asymmetric antenna configurations which supports not only communications between an access point and devices but also device-to-device (D2D communications for the Internet of things. For the network, we propose the transmit and receive beamforming with the channel state information (CSI for virtual multiple-input multiple-output (MIMO enabled by D2D receive cooperation. We analyze the sum rate achieved by a device pair in the proposed method and identify the strategies to improve the sum rate of the device pair. We next present a distributed algorithm and its equivalent algorithm for device pairing to maximize the throughput of the multidevice network. Simulation results confirm the advantages of the transmit CSI and D2D cooperation as well as the validity of the distributive algorithm.

Release strategies for making transferable semiconductor structures, devices and device components

Science.gov (United States)

Rogers, John A; Nuzzo, Ralph G; Meitl, Matthew; Ko, Heung Cho; Yoon, Jongseung; Menard, Etienne; Baca, Alfred J

2014-11-25

Provided are methods for making a device or device component by providing a multilayer structure having a plurality of functional layers and a plurality of release layers and releasing the functional layers from the multilayer structure by separating one or more of the release layers to generate a plurality of transferable structures. The transferable structures are printed onto a device substrate or device component supported by a device substrate. The methods and systems provide means for making high-quality and low-cost photovoltaic devices, transferable semiconductor structures, (opto-)electronic devices and device components.

Wireless device monitoring methods, wireless device monitoring systems, and articles of manufacture

Science.gov (United States)

McCown, Steven H [Rigby, ID; Derr, Kurt W [Idaho Falls, ID; Rohde, Kenneth W [Idaho Falls, ID

2012-05-08

Wireless device monitoring methods, wireless device monitoring systems, and articles of manufacture are described. According to one embodiment, a wireless device monitoring method includes accessing device configuration information of a wireless device present at a secure area, wherein the device configuration information comprises information regarding a configuration of the wireless device, accessing stored information corresponding to the wireless device, wherein the stored information comprises information regarding the configuration of the wireless device, comparing the device configuration information with the stored information, and indicating the wireless device as one of authorized and unauthorized for presence at the secure area using the comparing.

Class 1 devices case studies in medical devices design

CERN Document Server

Ogrodnik, Peter J

2014-01-01

The Case Studies in Medical Devices Design series consists of practical, applied case studies relating to medical device design in industry. These titles complement Ogrodnik's Medical Device Design and will assist engineers with applying the theory in practice. The case studies presented directly relate to Class I, Class IIa, Class IIb and Class III medical devices. Designers and companies who wish to extend their knowledge in a specific discipline related to their respective class of operation will find any or all of these titles a great addition to their library. Class 1 Devices is a companion text to Medical Devices Design: Innovation from Concept to Market. The intention of this book, and its sister books in the series, is to support the concepts presented in Medical Devices Design through case studies. In the context of this book the case studies consider Class I (EU) and 510(k) exempt (FDA) . This book covers classifications, the conceptual and embodiment phase, plus design from idea to PDS. These title...

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).

US plans for burning plasma experiments

International Nuclear Information System (INIS)

Nelson, D.

1982-01-01

The first US burning plasma experiment will be the TFTR at Princeton Plasma Physics Laboratory. The initial start-up with hydrogen is expected in December, 1983. The experiment by D-T reaction will begin in 1986. Because of the lack of shielding capability, later experiment is not yet defined. The informal scientific interaction with JET (European project) is kept. The design work on the Fusion Engineering Device (FED) continues, but is delayed. US fusion laboratories collaborated with IPP-Garching on the conceptual design of Zephyr experiment. The US continues to participate in INTOR activities, and will investigate into the critical issues relevant to both INTOR and FED in coming years. (Kato, T.)

The ORNL plasma fueling program

International Nuclear Information System (INIS)

Gouge, M.J.; Milora, S.L.; Foster, C.A.

1989-01-01

The Oak Ridge National Laboratory (ORNL) has been developing pellet injectors for over ten years. These devices produce frozen hydrogen isotope pellets and then accelerate the projectiles to speeds in the km/s range by either pneumatic or mechanical techniques. A variety of designs have been developed, including single-shot guns, multiple-shot guns, machine guns, and centrifugal accelerators. These injectors have been used to inject hydrogen and deuterium pellets into plasmas on numerous tokamak experiments resulting in improved plasma performance. ORNL has recently provided pellet fueling systems for the Tokamak Fusion Test Reactor (TFTR), the Joint European Torus (JET), and the Tore Supra tokamak. This paper discusses developments on these injector designs. 10 refs

The Burning Plasma Experiment conventional facilities

International Nuclear Information System (INIS)

Commander, J.C.

1991-01-01

The Burning Program Plasma Experiment (BPX) is phased to start construction of conventional facilities in July 1994, in conjunction with the conclusion of the Tokamak Fusion Test Reactor (TFTR) project. This paper deals with the conceptual design of the BPX Conventional Facilities, for which Functional and Operational Requirements (F ampersand ORs) were developed. Existing TFTR buildings and utilities will be adapted and used to satisfy the BPX Project F ampersand ORs to the maximum extent possible. However, new conventional facilities will be required to support the BPX project. These facilities include: The BPX building; Site improvements and utilities; the Field Coil Power Conversion (FCPC) building; the TFTR modifications; the Motor Generation (MG) building; Liquid Nitrogen (LN 2 ) building; and the associated Instrumentation and Control (I ampersand C) systems. The BPX building will provide for safe and efficient shielding, housing, operation, handling, maintenance and decontamination of the BPX and its support systems. Site improvements and utilities will feature a utility tunnel which will provide a space for utility services--including pulse power duct banks and liquid nitrogen coolant lines. The FCPC building will house eight additional power supplied for the Toroidal Field (TF) coils. The MG building will house the two MG sets larger than the existing TFTR MG sets. This paper also addresses the conventional facility cost estimating methodology and the rationale for the construction schedule developed. 6 figs., 1 tab

Neutral Beam Power System for TPX

International Nuclear Information System (INIS)

Ramakrishnan, S.; Bowen, O.N.; O'Conner, T.; Edwards, J.; Fromm, N.; Hatcher, R.; Newman, R.; Rossi, G.; Stevenson, T.; von Halle, A.

1993-01-01

The Tokamak Physics Experiment (TPX) will utilize to the maximum extent the existing Tokamak Fusion Test Reactor (TFTR) equipment and facilities. This is particularly true for the TFTR Neutral Beam (NB) system. Most of the NB hardware, plant facilities, auxiliary sub-systems, power systems, service infrastructure, and control systems can be used as is. The major changes in the NB hardware are driven by the new operating duty cycle. The TFTR Neutral Beam was designed for operation of the Sources for 2 seconds every 150 seconds. The TPX requires operation for 1000 seconds every 4500 seconds. During the Conceptual Design Phase of TPX every component of the TFTR NB Electrical Power System was analyzed to verify whether the equipment can meet the new operational requirements with our without modifications. The Power System converts 13.8 kV prime power to controlled pulsed power required at the NB sources. The major equipment involved are circuit breakers, auto and rectifier transformers surge suppression components, power tetrodes, HV Decks, and HVDC power transmission to sources. Thermal models were developed for the power transformers to simulate the new operational requirements. Heat runs were conducted for the power tetrodes to verify capability. Other components were analyzed to verify their thermal limitations. This paper describes the details of the evaluation and redesign of the electrical power system components to meet the TPX operational requirements

PLASMA DEVICE

Science.gov (United States)

Gow, J.D.; Wilcox, J.M.

1961-12-26

A device is designed for producing and confining highenergy plasma from which neutrons are generated in copious quantities. A rotating sheath of electrons is established in a radial electric field and axial magnetic field produced within the device. The electron sheath serves as a strong ionizing medium to gas introdueed thereto and also functions as an extremely effective heating mechanism to the resulting plasma. In addition, improved confinement of the plasma is obtained by ring magnetic mirror fields produced at the ends of the device. Such ring mirror fields are defined by the magnetic field lines at the ends of the device diverging radially outward from the axis of the device and thereafter converging at spatial annular surfaces disposed concentrically thereabout. (AFC)

Practical microwave electron devices

CERN Document Server

Meurant, Gerard

2013-01-01

Practical Microwave Electron Devices provides an understanding of microwave electron devices and their applications. All areas of microwave electron devices are covered. These include microwave solid-state devices, including popular microwave transistors and both passive and active diodes; quantum electron devices; thermionic devices (including relativistic thermionic devices); and ferrimagnetic electron devices. The design of each of these devices is discussed as well as their applications, including oscillation, amplification, switching, modulation, demodulation, and parametric interactions.

Medical Devices; General Hospital and Personal Use Devices; Classification of the Ultraviolet Radiation Chamber Disinfection Device. Final order.

Science.gov (United States)

2015-11-20

The Food and Drug Administration (FDA or the Agency) is classifying the ultraviolet (UV) radiation chamber disinfection device into class II (special controls). The special controls that will apply to the device are identified in this order and will be part of the codified language for the UV radiation chamber disinfection device classification. The Agency is classifying the device into class II (special controls) in order to provide a reasonable assurance of safety and effectiveness of the device.

Photovoltaic device

Energy Technology Data Exchange (ETDEWEB)

Reese, Jason A; Keenihan, James R; Gaston, Ryan S; Kauffmann, Keith L; Langmaid, Joseph A; Lopez, Leonardo; Maak, Kevin D; Mills, Michael E; Ramesh, Narayan; Teli, Samar R

2017-03-21

The present invention is premised upon an improved photovoltaic device ("PV device"), more particularly to an improved photovoltaic device with a multilayered photovoltaic cell assembly and a body portion joined at an interface region and including an intermediate layer, at least one interconnecting structural member, relieving feature, unique component geometry, or any combination thereof.

Photovoltaic device

Science.gov (United States)

Reese, Jason A.; Keenihan, James R.; Gaston, Ryan S.; Kauffmann, Keith L.; Langmaid, Joseph A.; Lopez, Leonardo C.; Maak, Kevin D.; Mills, Michael E.; Ramesh, Narayan; Teli, Samar R.

2015-06-02

The present invention is premised upon an improved photovoltaic device ("PV device"), more particularly to an improved photovoltaic device with a multilayered photovoltaic cell assembly and a body portion joined at an interface region and including an intermediate layer, at least one interconnecting structural member, relieving feature, unique component geometry, or any combination thereof.

Residual effects of chromium gettering on the outgassing behavior of a stainless steel vacuum vessel

International Nuclear Information System (INIS)

Simpkins, J.E.; Blanchard, W.R.; Dylla, H.F.; LaMarche, P.H.

1986-05-01

Laboratory experiments that compared chromium and titanium gettering showed that with chromium, unlike titanium, there is no appreciable diffusion of hydrogen isotopes into the film. It was concluded from these experiments that chromium gettering on tokamaks is more desirable than titanium gettering, since chromium should provide higher hydrogen recycling, minimize tritium inventories, and avoid hydrogen embrittlement. Large-scale sublimation sources, consisting of hollow elongated chromium spheres with internal resistance heaters, were developed for use on tokamaks. These sources have been used to getter both the Impurity Study Experiment (ISX) and the Tokamak Fusion Test Reactor (TFTR). In both bases, significant effects on plasma performance were observed, including lower Z/sub eff/ and radiated power losses and an increase in the density limit. In TFTR these effects were observed for a period of weeks after a single chromium deposition. This paper reports the results of laboratory experiments made to examine the gettering characteristics of chromium films under conditions simulating those in TFTR

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.)

High voltage semiconductor devices and methods of making the devices

Energy Technology Data Exchange (ETDEWEB)

Matocha, Kevin; Chatty, Kiran; Banerjee, Sujit

2018-01-23

A multi-cell MOSFET device including a MOSFET cell with an integrated Schottky diode is provided. The MOSFET includes n-type source regions formed in p-type well regions which are formed in an n-type drift layer. A p-type body contact region is formed on the periphery of the MOSFET. The source metallization of the device forms a Schottky contact with an n-type semiconductor region adjacent the p-type body contact region of the device. Vias can be formed through a dielectric material covering the source ohmic contacts and/or Schottky region of the device and the source metallization can be formed in the vias. The n-type semiconductor region forming the Schottky contact and/or the n-type source regions can be a single continuous region or a plurality of discontinuous regions alternating with discontinuous p-type body contact regions. The device can be a SiC device. Methods of making the device are also provided.

Process control device

International Nuclear Information System (INIS)

Hayashi, Toshifumi; Kobayashi, Hiroshi.

1994-01-01

A process control device comprises a memory device for memorizing a plant operation target, a plant state or a state of equipments related with each other as control data, a read-only memory device for storing programs, a plant instrumentation control device or other process control devices, an input/output device for performing input/output with an operator, and a processing device which conducts processing in accordance with the program and sends a control demand or a display demand to the input/output device. The program reads out control data relative to a predetermined operation target, compares and verify them with actual values to read out control data to be a practice premise condition which is further to be a practice premise condition if necessary, thereby automatically controlling the plant or requiring or displaying input. Practice presuming conditions for the operation target can be examined succesively in accordance with the program without constituting complicated logical figures and AND/OR graphs. (N.H.)

Fusion engineering. Vol. 2

International Nuclear Information System (INIS)

Young, N.E.; Pinter, G.R.; Spinos, F.R.

1983-01-01

An x-ray crystal spectrometer is scheduled for installation in the Tokamak Fusion Test Reactor in 1984 coinciding with the TFTR deuterium operation phase. This spectrometer is designed to measure the spectra of hydrogen-like and helium-like impurities in the plasma. Ion temperatures electron temperatures, electron density and the distribution of impurity charge states are obtained from the x-ray detector count ratios. The velocity of the toroidal rotation of the plasma is also discerned using this device. The x-ray crystal spectrometer is based on the Bragg diffraction of x-rays from a curved crystal impinging on a multiwire proportional counter. Only those x-rays that satisfy the Bragg relationship (lambda =2d sin theta) will be diffracted and strike the proportional counter. This paper is limited to a discussion of the physical characteristics of the spectrometer and the methods devised to satisfy the operational aspects of such a device

Repeating pneumatic hydrogen pellet injector for plasma fueling

International Nuclear Information System (INIS)

Combs, S.K.; Milora, S.L.; Foust, C.R.; Foster, C.A.; Schuresko, D.D.

1985-01-01

A repeating pneumatic pellet injector has been developed for plasma fueling applications. The repetitive device extends pneumatic injector operation to steady state. The active mechanism consists of an extruder and a gun assembly that are cooled by flowing liquid-helium refrigerant. The extruder provides a continuous supply of solid hydrogen to the gun assembly, where a reciprocating gun barrel forms and chambers cylindrical pellet from the extrusion; pellets are then accelerated with compressed hydrogen gas (pressures up to 125 bar) to velocities -1 have been obtained with 2.1- , 3.4- , and 4.0-mm-diameter pellets. The present apparatus operates at higher firing rates in short bursts; for example, a rate of 6 s -1 for 2 s with the larger pellets. These pellet parameters are in the range applicable for fueling large present-day fusion devices such as the Tokamak Fusion Test Reactor (TFTR). Experimental results are presented, including effects of propellant pressure and barrel length on gun performance

Pellet fueling development at ORNL

International Nuclear Information System (INIS)

Combs, S.K.; Milora, S.L.; Foster, C.A.; Schuresko, D.D.; Foust, C.R.; Simmons, D.W.; Beard, D.S.

1986-09-01

Advanced plasma fueling systems for magnetic confinement devices are being developed at the Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogenic pellets at speeds in the range of 1-2 km/s and higher. Two specific concepts are under development: (1) high-speed pneumatic acceleration; and (2) mechanical (centrifugal) acceleration. Both approaches are being pursued to meet the projected pellet size and delivery rates for major near-term plasma confinement devices, such as the Tokamak Fusion Test Reactor (TFTR), Tore Supra, the Joint European Torus (JET), JT-60, and Doublet III-D (DIII-D), as well as future applications. In addition to these confinement physics related activities, ORNL is pursuing advanced technologies to achieve pellet velocities significantly in excess of the 2-km/s range already attained with pneumatic injectors and has embarked on a development program designed to explore the feasibility of fabricating and accelerating tritium pellets. This paper describes these ongoing activities

Hip supporting device

DEFF Research Database (Denmark)

2011-01-01

The present invention relates to a device for limiting movements in one or more anatomical joints, such as a device for limiting movement in the human hip joint after hip replacement surgery. This is provided by a device for limiting movement in the human hip joint, said device comprising: at least...

Cooling device in thermonuclear device

International Nuclear Information System (INIS)

Honda, Tsutomu.

1988-01-01

Purpose: To prevent loss of cooling effect over the entire torus structure directly after accidental toubles in a cooling device of a thermonuclear device. Constitution: Coolant recycling means of a cooling device comprises two systems, which are alternately connected with in-flow pipeways and exit pipeways of adjacent modules. The modules are cooled by way of the in-flow pipeways and the exist pipeways connected to the respective modules by means of the coolant recycling means corresponding to the respective modules. So long as one of the coolant recycling means is kept operative, since every one other modules of the torus structure is still kept cooled, the heat generated from the module put therebetween, for which the coolant recycling is interrupted, is removed by means of heat conduction or radiation from the module for which the cooling is kept continued. No back-up emergency cooling system is required and it can provide high economic reliability. (Kamimura, M.)

Photovoltaic device

Science.gov (United States)

Reese, Jason A.; Keenihan, James R.; Gaston, Ryan S.; Kauffmann, Keith L.; Langmaid, Joseph A.; Lopez, Leonardo C.; Maak, Kevin D.; Mills, Michael E.; Ramesh, Narayan; Teli, Samar R.

2015-09-01

The present invention is premised upon an improved photovoltaic device ("PV device"), more particularly to an improved photovoltaic device (10) with a multilayered photovoltaic cell assembly (100) and a body portion (200) joined at an interface region (410) and including an intermediate layer (500), at least one interconnecting structural member (1500), relieving feature (2500), unique component geometry, or any combination thereof.

78 FR 68714 - Medical Devices; Ophthalmic Devices; Classification of the Scleral Plug

Science.gov (United States)

2013-11-15

... amendments), as ``preamendments devices.'' FDA classifies these devices after the Agency takes the following.... FDA-2012-N-1238] Medical Devices; Ophthalmic Devices; Classification of the Scleral Plug AGENCY: Food... scleral plugs in order to provide a reasonable assurance of safety and effectiveness of the device. The...

Hardware design of a microcomputer controlled diagnostic vacuum controller

International Nuclear Information System (INIS)

Marsala, R.J.

1983-01-01

The TFTR diagnostic vacuum controller (DVC) has been designed and built to control and monitor the pumps, valves and gauges which comprise a diagnostic vacuum system. The DVC is a microcomputer based self-contained controller with battery backup which may be controlled manually from front panel controls or remotely via CICADA. The DVC implements all pump and valve sequencing and provides protection against incorrect operation. There are presently two versions of the DVC operating on TFTR and a third version being used on the S-1 machine

21 CFR 864.9195 - Blood mixing devices and blood weighing devices.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Blood mixing devices and blood weighing devices. 864.9195 Section 864.9195 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES HEMATOLOGY AND PATHOLOGY DEVICES Products Used In Establishments That...