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The DEGAS neutral transport code is used in two separate cases to simulate the neutral beam box and vessel of the Tokamak Fusion Test Reactor (TFTR). For the neutral beam box simulation, known input parameters include the ion density at the source exit and the proportion of input gas that is converted to the high-energy atomic beam. The T 0 current to the torus is (1.61 ± 0.03) X 10 20 s -1 , with the high-energy beam having a median energy above 95 keV. Corresponding results are found for the D 0 current. In addition, the amount of gas reaching the torus, the pressure, and the flux and energy distributions of the ions and neutrals to the walls are found. For the tritium case, it is calculated that 92.4 ± 0.2% of the input tritium reaches the cryopanels, 6.64 ± 0.05% reaches the torus, and 1.0 ± 0.2% reaches the ion dump. In the second run, DEGAS was used to calculate the neutral atom flux and energy ofparticles incident on the walls of the vacuum vessel and the neutral pressure in the pump duct of TFTR during a typical supershot with a 50/50 mixture of deuterium-tritium. Output quantities are the current and energy to the bumper limiter and first wall. The total amount of tritium implanted in the vacuum vessel after 150 shots of 1-s duration is estimated to be 0.5 ± 0.1 g in the bumper limiter and 0.042 ± 0.023 g in the outer wall and pumping duct, which is well within the 5-g on-site inventory and the 2-g in-vessel inventory. The implications of these results are discussed.

A first set of shock propagation, laser-plasma interaction, hohlraum energetics and hydrodynamic experiments have been performed using the first 4 beams of the National Ignition Facility (NIF), in support of indirect drive Inertial Confinement Fusion (ICF) and High Energy Density Physics.

Tritium Balmer-alpha (T{sub {alpha}}) emission, along with H{sub {alpha}} and D{sub {alpha}} 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{sub {alpha}} is relatively slow to appear in tritium neutral beam heated discharges, (T{sub {alpha}}/(H{sub {alpha}} + D{sub {alpha}} + T{sub {alpha}}) = 11% after 8 tritium-only neutral beam discharges). In contrast, the T{sub {alpha}} fraction in a sequence of six discharges fueled with tritium puff,s increased to 44%. Larger transient increases (up to 75% T{sub {alpha}}) 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.

Operation of the Tokamak Fusion Test Reactor (TFTR) with a mixture of deuterium and tritium fueling has permitted the opportunity to measure the retention of tritium in the graphite limiter and other internal hardware. The use of discharge cleaning techniques and venting to remove the tritium was investigated. The tritium was introduced into TFTR by neutral beam injection and by gas puffing. The graphite limiter is subject to erosion and codeposition. While short term retention was high, the retention averaged over the 1993-1995 D-T campaign was 52% {+-} 15%. The tritium removal techniques resulted in lowering the in-vessel inventory from 16.4 kCi at the end of 1995 operation to 7.2 kCi at the start of the 1996 experimental program. 14 refs., 3 figs., 1 tab.

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.

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma.The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral‐beam heating, in a supershot discharge and 6.7 MW in a high‐βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter‐H‐mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high‐βp discharges. Ion cyclotron range of frequencies (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 [Nucl. Fusion 34, 1247 (1994)] 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. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

TFTR has begun its campaign to study deuterium-tritium fusion under reactor-like conditions. Variable amounts of deuterium and tritium neutral beam power have been used to maximize fusion power, study alpha heating, investigate alpha particle confinement, and search for alpha driven plasma instabilities. Additional areas of study include energy and particle transport and confinement, ICRF heating schemes for DT plasmas, tritium retention, and fusion in high βp plasmas. The majority of this work is done in the TFTR supershot confinement regime. To obtain supershots, extensive limiter conditioning using helium fueled ohmic discharges and lithium pellet injection into ohmic and neutral beam heated plasmas is performed, resulting in a low recycling limiter. The relationship between recycling and core plasma confinement has been studied by using helium, deuterium and high-Z gas puffs to simulate high recycling limiter conditions. These studies show that confinement in TFTR supershots is very sensitive to the influx of neutral particles at the plasma edge.

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 107 kCi 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 9 MW 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.

The first spectroscopic measurements of tritium Balmer α(Tα) emission from a fusion plasma were made on TFTR using a Fabry-Perot interferometer. The tritium α(Tα) emission line is partially blended with the deuterium α(Dα) line, commonly used in edge plasma diagnostics, and the contributions of Hα, Dα and Tα are separated by spectral analysis. The data are a measure of the fuelling of the plasma by tritium accumulated in the TFTR limiter, as well as the amount of neutral tritium generated by charge exchange of plasma ions. The Tα line first became detectable in a high power, tritium only, neutral beam injection discharge at the level of Tα/(Hα + Dα + Tα) = 2%. Subsequently this ratio has increased to as high as 7.5%. Data on the time evolution of the Tα emission during a single discharge and over a series of tritium and deuterium discharges are presented

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

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a FluorinertTM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium–deuterium (D–D) run to simulate expected deuterium–tritium (D–T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D–T experiments using D–D have been performed. The physics objectives of D–T operation are production of ≊10 MW of fusion power, evaluation of confinement, and heating in deuterium–tritium plasmas, evaluation of α‐particle heating of electrons, and collective effects driven by alpha particles and testing of diag...

Recent research on TFTR has emphasized optimization of performance in deuterium plasmas, transport studies and studies of energetic ion and fusion product physics in preparation for the D-T experiments that will commence in July of 1993. TFTR has achieved full hardware design parameters, and the best TFTR discharges in deuterium are projected to QDT of 0.3 to 0.5.The physics phenomena that will be studied during the D-T phase will include: tritium particle confinement and fueling, ICRF heating with tritium, species scaling with tritium, collective alpha-particle instabilities, alpha heating of the plasma and helium ash buildup. It is important for the fusion program that these physics issues be addressed to identify regimes of benign alpha behavior, and to develop techniques to actively stabilize or control instabilities driver by collective alpha effects.

Beam diagnostic instrumentation is being developed for the LEDA, a 6.7-MeV, 100-mA-cw proton accelerator, presently being commissioned at the Los Alamos National Laboratory (LANL). This instrumentation will be the basis for much of the Accelerator Production of Tritium and the Spallation Neutron Source linac. Located in the LEDA injector and the high energy beam transport (HEBT) this initial instrumentation suite's purpose is to verify the RFQ pulsed and cw operation. The instrumentation include a series of DC, pulsed- and bunched-beam current measurements from which RFQ beam-transmission efficiency will be determined. Ionization-chamber beam loss measurements are mounted above the HEBT and provide input signals to a fast equipment protection system. Central beam phase and energy measurements provide RFQ longitudinal performance information. Beam position measurements provide information to properly center the beam within the HEBT beam pipe. Finally, two types of transverse profile measurements including a slow wire scanner and a video fluorescence monitor provide beam width and projection information in the LEDA HEBT. This paper will discuss these measurements developed for LEDA and summarize how they performed during RFQ verification experiments.

A 52 quadrupole-magnet FODO lattice has been assembled and operated at the Los Alamos National Laboratory. The purpose of this lattice is to provide a platform to measure the resulting beam halo as the first four magnets of the lattice produce various mismatch conditions. These data are then compared with particle simulations so that halo formation mechanisms may be better understood. The lattice is appended to the LEDA 6.7-MeV radiofrequency quadrupole (RFQ) and is followed by a short high-energy beam transport (HEBT) that safely dumps the beam into a 670-kW beam stop. Beam diagnostic instruments are interspersed within the lattice and HEBT. The primary instruments for measuring the beam halo are nine interceptive devices that acquire the beam's horizontal and vertical projected particle density distributions out to an approximate 10/sup 5/:1 dynamic range. These distributions are acquired using both traditional wire scanners and water-cooled graphite scraping devices. The lattice and HEBT instrumentation set also includes position, bunched-beam current, pulsed current, and beam loss measurements. This paper briefly describes and details the operation of each instrument compares measured data from the different types of instruments, and refers to other detailed papers.

In late 1993 the Tokamak Fusion Test Reactor began operating with a deuterium-tritium (DT) fuel mixture instead of the pure deuterium which it had used previously. The major portion of this tritium has initially entered the torus as energetic neutral beam particles. Over 600 deuterium-tritium discharges have now been studied with the aid of more than 2000 tritium ion source shots. The maximum total neutral particle power injected with a mix of deuterium and tritium beams has been 39.6 megawatts, and the maximum injected as tritium neutrals has been 24.3 megawatts. Tritium neutral beam operation has become routine during this time.

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 {approx} 8,000 Ci and restoring the tritium inventory to a level well below the administrative limit.

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%.

Observation of the intensity of the recycling particle flux at the main plasma edge for various limiter and divertor discharges indicates that the gross energy confinement of beam-heated discharges is closely related to the intensity of the edge particle flux. In limiter discharges, the global particle confinement time and the energy confinement time τE show many similarities: 1) linear Ip dependence at Ip < 600 kA, 2) no BT dependence, and 3) deterioration against injection power. Improvement of τE by increasing Ip, for example, is associated with high temperatures at the plasma edge region accompanied by reduced particle recycling. – Divertor discharges with low particle recycling around the main plasma show better energy confinement than limiter discharges at high plasma densities. The improvement of τE is primarily originated in the reduction of heat transport at the main plasma edge region, which is associated with the reduction of recycling particle flux at the main plasma edge. Under certain operation condition, for example, excessive cold-gas puffing, the discharge shows relatively high scrape-off plasma density and strong particle recycling between the main plasma and the limiter. The energy confinement time of these discharges degrades somewhat or reduces completely to that of the limiter discharge. – In low-recycling divertor discharges, the central electron and ion temperature is proportional to the injection power, and the plasma stored energy is proportional to ePabs (scales as INTOR scaling). With ≈ 4 MW beam injection, high-temperature and high-density plasmas were obtained (stored energy up to 280 kJ, Te(0) ≈ Ti(0) ≈ 2.5–3.0 keV at e ≈ (6–7) × 1013 cm−3, τE* ≈ 70 ms).

The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory is considering the use of tritium neutral beams during the deuterium–tritium phase of operation in 1990. A concern is the tritium inventory that will develop in neutral beam components such as the copper calorimeter. In this paper we report on measurements and calculations of hydrogen isotope retention and release for copper under ion bombardment. Retention of deuterium trapped in the near‐surface region is measured as a function of temperature using the D(3He, p) α nuclear reaction technique with an 800‐keV 3He+ analysis beam. The rate of molecular recombination (Kr) at the copper–vacuum interface is determined from plasma driven permeation studies in our Tritium Plasma Experiment facility. The measurements indicate that Kr=6.7±0.1×10−18 cm4/s over the temperature range 575–825 K, in agreement with theory. These data are used as inputs to the diffuse computer code to provide a model for tritium–copper interactions. The mode...

Experiments with plasmas having nearly equal concentrations of deuterium and tritium have been carried out on TFTR. To date, the maximum fusion power has been 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-{beta}{sub p} discharge following a current ramp-down. The fusion power density in the core of the plasma has reached 2.8 MWm{sup {minus}3}, exceeding that expected in the International Thermonuclear Experimental Reactor (ITTER). The energy confinement time, {tau}{sub E}, is observed to increase in D-T, relative to D plasmas, by 20% and the n{sub i}(O){center_dot}{tau}{sub E} product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-{beta}{sub p} discharges. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations assuming classical confinement. Measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from helium gas puffing experiments. The loss of energetic alpha particles to a detector at the bottom ofmore » the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. ICRF heating of a D-T plasma, using the second harmonic of tritium, has been demonstrated. D-T experiments on TFTR will continue both to explore the physics underlying the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.« less