Publications

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Overview of Plasma Materials Interaction (PMI) activities are: (1) Hydrogen diffusion and trapping in metals - (a) Growth of hydrogen precipitates in tungsten PFCs, (b) Temperature dependence of deuterium retention at displacement damage, (c) D retention in W at elevated temperatures; (2) Permeation - (a) Gas driven permeation results for W/Mo/SiC, (b) Plasma-driven permeation test stand for TPE; and (3) Surface studies - (a) H-sensor development, (b) Adsorption of oxygen and hydrogen on beryllium surfaces.

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.

The mechanisms governing D loss in ErD2 films with and without a Mo diffusion barrier on kovar substrates were studied between 200 and 600 °C via in-situ Ion Beam Analysis (IBA). Significant intermixing between kovar and Er was observed above 450°C and between kovar and ErD2 above 500 °C. The D loss mechanism in ErD2 films was found to change from intermixing between kovar and ErD2 at low temperatures (< 500 °C) to thermal decomposition at higher temperatures (> 500 °C). Diffusion between kovar and ErD2 was measured isothermally at 500 and 550 °C. An activation energy of 2.1 eV and a pre-exponential factor of 0.071 cm2/s were determined. Diffusion between the kovar components and ErD2 film was inhibited by depositing a 200 nm Mo diffusion barrier between the kovar substrate and the ErD2 film. The processing of the Mo diffusion barrier was shown to impact its performance. Intermixing between the kovar / Mo / ErD2 stack becomes significant between 500 and 550 °C with a sputter deposited Mo diffusion barrier and between 550 and 600 °C for an electron-beam evaporated Mo diffusion barrier. Copyright © 2009 ASM International® All rights reserved.

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{sm_bullet}Mixing from some thermal process steps thought to drive H,D,T loss - This does not appear to be a problem with the Mo/Er occluder stacks {sm_bullet}Diffusion barriers investigated to prevent mixing

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Deposition in next-step devices such as ITER will pose diagnostic challenges. Codeposition of hydrogen with carbon needs to be characterized and understood in the initial hydrogen phase in order to mitigate tritium retention and qualify carbon plasma facing components for DT operations. Plasma facing diagnostic mirrors will experience deposition that is expected to rapidly degrade their reflectivity, posing a challenge to diagnostic design. Some eroded particles will collect as dust on interior surfaces and the quantity of dust will be strictly regulated for safety reasons however, diagnostics of in-vessel dust are lacking. We report results from two diagnostics that relate to these issues. Measurements of deposition on NSTX with 4 Hz time resolution have been made using a quartz microbalance in a configuration that mimics that of a typical diagnostic mirror. Often deposition was observed immediately following the discharge suggesting that diagnostic shutters should be closed as soon as possible after the time period of interest. Material loss was observed following a few discharges. A novel diagnostic to detect dust particles on remote surfaces was commissioned on NSTX.

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7 MW) and high harmonic fast wave (HHFW) heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with {beta}{sub T} {triple_bond} <p>/(B{sub T0}{sup 2}/2{mu}{sub 0}) of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fueling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparison of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10 MW m{sup -2} has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta ({beta}), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values {beta}{sub T} of up to 35% with a near unity central {beta}{sub T} have been obtained. NSTX will be exploring advanced regimes where {beta}{sub T} up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction ({approx}60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to I{sub p} {approx} 500 kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

{sup 13}CH{sub 4} was injected with a toroidally-symmetric gas system into 22 identical lower-single-null L-mode discharges on DIII-D. The injection level was adjusted so that it did not significantly perturb the core or divertor plasmas, with a duration of {approx}3 s on each shot, for a total of {approx}300 T L of injected particles. The plasma shape remained very constant; the divertor strike points were controlled to {approx}1 cm at the divertor plate. At the beginning of the subsequent machine vent, 29 carbon tiles were removed for nuclear reaction analysis of {sup 13}C content to determine regions of carbon deposition. It was found that only the tiles inboard of the inner strike point had appreciable {sup 13}C above background. Visible spectroscopy measurements of the carbon injection and comparisons with modeling are consistent with carbon transport by means of scrape-off layer flow.

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.

The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last international atomic energy agency (IAEA) meeting, we have made significant progress in developing the building blocks needed for AT operation: (1) we have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, we have achieved {beta}{sub N}H{sub 89} {ge} 10 for 4{tau}{sub E} limited by the neoclassical tearing mode (NTM); (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, we have stabilized the (m, n) = (3, 2) NTM and then increased {beta}{sub T} by 60%; (4) we have produced ECCD stabilization of the (2, 1) NTM in initial experiments; (5) we have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) we have demonstrated stationary tokamak operation for 6.5 s (36{tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx_equal} as ITER but at much higher q{sub 95} = 4.2. We have developed general improvements applicable to conventional and AT operating modes: (1) we have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, edge localized modes (ELM) heat load to the divertor and which can run for long periods of time (3.8 s or 25{tau}{sub E}) with constant density and constant radiated power; (2) we have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; (3) we have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much smaller than to the outer strike points. We have made detailed investigations of the edge pedestal and scrape-off layer (SOL): (1) atomic physics and plasma physics both play significant roles in setting the width of the edge density barrier in H-mode; (2) ELM heat flux conducted to the divertor decreases as density increases; (3) intermittent, bursty transport contributes to cross field particle transport in the SOL of H-mode and, especially, L-mode plasmas.