A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.
Unbalanced double null ELMy H-mode configurations in DIII-D are used to simulate the situation in ITER high triangularity, burning plasma magnetic equilibria, where the second X-point lies close to the top of the vacuum vessel, creating a secondary divertor region at the upper blanket modules. The measured plasma conditions in the outer secondary divertor closely duplicated those projected for ITER. {sup 13}CH{sub 4} was injected into the secondary outer divertor to simulate sputtering there. The majority of the {sup 13}C found was in the secondary outer divertor. This material migration pattern is radically different than that observed for main wall {sup 13}CH{sub 4} injections into single null configurations where the deposition is primarily at the inner divertor. The implications for tritium codeposition resulting from sputtering at the secondary divertor in ITER are significant since release of tritium from Be co-deposits at the main wall bake temperature for ITER, 240 C, is incomplete. The principal features of the measured {sup 13}C deposition pattern have been replicated by the OEDGE interpretive code.
As part of a study of carbon-tritium co-deposition, we carried out an experiment on DIII-D involving a toroidally symmetric injection of {sup 13}CH{sub 4} at the top of a LSN discharge. A Monte Carlo code, DIVIMP-HC, which includes molecular breakup of hydrocarbons, was used to model the region near the puff. The interpretive analysis indicates a parallel flow in the SOL of M {parallel} {approx} 0.4 directed toward the inner divertor. The CH{sub 4} is ionized in the periphery of the SOL and so the particle confinement time, T{sub C}, is not high, only {approx} 5 ms, and about 4X lower than if the CH{sub 4} were ionized at the separatrix. For such a wall injection location, however, approximately 60-75% of the CH{sub 4} gets ionized to C{sup +}, C{sup 2+}, etc., and is efficiently transported along the SOL to the inner divertor, trapping hydrogen by co-deposition there.
A new reciprocating Langmuir probe has been used to measure density and temperature profiles, ion flow, and potential fluctuation levels from the lower divertor floor up to the X-point on the DIII-D tokamak. This probe is designed to make fast (2 kHz swept, 20 kHz Mach, 500 kHz Vfloat) measurements with 2 mm spatial resolution in the region where the largest gradients on the plasma open flux tubes are found and therefore provide the best benchmarks for SOL and divertor numerical models. Profiles are constructed using the 300 ms time history of the probe measurements during the 25 cm reciprocating stroke. Both single and double null plasmas can be measured and compared with a 20 Hz divertor Thomson scattering system. The probe head is constructed of four different kinds of graphite to optimize the electrical and thermal characteristics. Electrically insulated pyrolytic graphite rings act as a heat shield to absorb the plasma heat flux on the probe shaft and are mounted on a carbon/carbon composite core for mechanical strength. The Langmuir probe sampling tips are made of a linear carbon fiber composite. The mechanical, electrical, data acquisition and power supply systems design will be described. Initial measurements will also be presented.
This paper describes divertor density and temperature measurements using both a new reciprocating Langmuir probe (XPT-RCP) which plunges vertically above the divertor floor up to the X-point height and swept, single, Langmuir probes fixed horizontally across the divertor floor. These types of measurements are important for testing models of the SOL and divertor which then are used to design plasma facing components in reactor size tokamaks. This paper presents an overview of the new divertor probe measurements and how they compare with the new divertor Thomson scattering system. The fast time response of the probe measurements allows detailed study of ELMs.
Power deposition measurements have been carried out on the ALT-II toroidal belt pump limiter and the inner bumper limiter in TEXTOR for Ohmic, neutral beam and RF heated discharges. Two infrared cameras and the ALT-II thermocouple array indicate that {lambda}{sub E} remains unchanged (7 mm) in the presence of beams but increases to 10 mm with ICRH. The heating distribution is less uniform on the bumper limiter than on ALT-II, which potentially could explain the differences seen in graphite surface pumping. 9 refs., 3 figs., 1 tab.
The operation of TORE SUPRA at full power (25MW, 30s) has led to the design of a full set of actively pumped carbon limiters to remove at least 8MW and to partially control the particle balance. An interim version is now installed, composed of 5 vertical and one horizontal outboard (OPL) pump limiters, semi-inertially water cooled. The latter is a result of a collaboration between the US DOE and the Association EUR-CEA, it is fully instrumented and therefore can serve as a reference for the final design. Ohmic discharges (1.85T, 740kA, 8.5s) in helium have been used to test the thermal load on and the particle exhaust efficiency of the OPL. In these experiments the plasma is formed on the inner wall (R = 232 cm, a = 76 cm) and subsequently displaced (6 cm) outward, early on the current plateau, to lean on the OPL (R = 238 cm, a = 75 cm). In addition to the limiters above, a non-pumped outboard (ONLP) limiter of identical shape to the OPL served to produce similar discharges for better comparison and determination of particle control. A comparison is made hereafter of the thermal load and particle pumping effects on the OPL when the plasma is in contact either with the OPL/ONPL alone or with the OPL and the vertical limiters together. 3 refs., 1 fig., 2 tabs.
Particle collection, removal, and exhaust by the toroidal belt pump limiter ALT-III have been measured in deuterium discharges with co-, counter-, and balanced injection of 48 keV neutral hydrogen particles. Particle collection increases from 50-80 A to 150-320 A during 1.2 MW of co- or counter-injection or 2.4 MW of balanced injection. The removal rate for pumping at two of the eight blades (3 of 15 scoops) reaches 2.7 Torr-l/s with a removal efficiency of nearly 45%. Extrapolating these results to a full belt with 15 scoops and eight pumps yields 140 amps of removal. This compares favorably with the maximum injectable current of 50 A and suggests that ALT-II with full pumping can provide sufficient exhaust during NI heating. 4 figs.
It has been observed on TEXTOR that in low density discharges the electrons gain enough energy to emit relativistic synchrotron radiation in the 3--6 {mu}m IR-range, and this radiation is due to electrons with energies up to 30 MeV. The momentum in perpendicular direction amounts to about 1/10 of the longitudinal one. The total number of runaways is of the order of 10{sup 16} electrons, and they carry a current of about 20% of the total plasma current. 3 refs., 1 fig.