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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.

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.

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

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.

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).

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