Showing papers by "Richard E. Nygren published in 2009"

Plasma-surface interaction issues of an all-metal ITER

Abstract: We assess key plasma surface interaction issues of an all-metal plasma facing component (PFC) system for ITER, in particular a tungsten divertor surface, and a beryllium or tungsten first wall. Such a system eliminates problems with carbon divertor erosion and T/C codeposition, and for an all-tungsten system would better extrapolate to post-ITER devices. The issues studied are sputtering, transport, and formation of mixed surface layers, tritium codeposition, core plasma contamination, ELM response, and He on W irradiation effects. Code package OMEGA computes PFC sputtering erosion/redeposition in an ITER full power D-T plasma edge regime with convective transport. The HEIGHTS package analyzes divertor plasma transient response. PISCES and other data are used with code results to assess PFC performance. Predicted outer wall sputter erosion rates are acceptable for Be (0.3 nm/s) or bare (stainless steel/Fe) wall (0.05 nm/s) for the low duty factor ITER, and are very low (0.002 nm/s) for W. Most wall-sputtered Be is redeposited on the wall itself or baffle region, with about 10% transported to the divertor target. T/Be codeposition in redeposited wall material could be significant (~2 gT per 400 s ITER pulse). Core plasma contamination potential from wall sputtering appears acceptable for Be (~2%), and negligible for W (or Fe) due to near-surface ionization of sputtered W (Fe) atoms and subsequent strong redeposition. A tungsten divertor likewise appears acceptable from the self-sputtering and plasma contamination standpoints, and would have negligible T/W codeposition. Be can grow on/near the strike point region of a W divertor, but for the predicted maximum surface temperature of ~800°C, deleterious Be/W alloy formation may be avoided. ELM's are a serious challenge to the divertor, but this is true for all materials. We identify acceptable ELM parameters for W. We conclude that an all-metal PFC system is likely a much better choice for ITER D-T operation than a system using carbon, but critical R&D issues remain, e.g., in areas of transient surface erosion (of all materials), W surface integrity with energetic He etc. bombardment, and in predictive plasma/surface interaction modeling generally. Steps are suggested to ameliorate problems and reduce uncertainties, e.g., via a 300 or 400°C baking capability for T/Be reduction, and using a deposited tungsten first wall test section.

Overview of results from the National Spherical Torus Experiment (NSTX)

Abstract: The mission of the National Spherical Torus Experiment (NSTX) is the demonstration of the physics basis required to extrapolate to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high β operation. To better understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale fluctuations with varying electron temperature gradient scale length. Results from n = 3 braking studies are consistent with the flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved coupling of high harmonic fast-waves has been achieved by reducing the edge density relative to the critical density for surface wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation. Plasmas have been maintained on NSTX at very low internal inductance li ~ 0.4 with strong shaping (κ ~ 2.7, δ ~ 0.8) with βN approaching the with-wall β-limit for several energy confinement times. By operating at lower collisionality in this regime, NSTX has achieved record non-inductive current drive fraction fNI ~ 71%. Instabilities driven by super-Alfvenic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvenic. Linear toroidal Alfven eigenmode thresholds and appreciable fast ion loss during multi-mode bursts are measured and these results are compared with theory. The impact of n > 1 error fields on stability is an important result for ITER. Resistive wall mode/resonant field amplification feedback combined with n = 3 error field control was used on NSTX to maintain plasma rotation with β above the no-wall limit. Other highlights are results of lithium coating experiments, momentum confinement studies, scrape-off layer width scaling, demonstration of divertor heat load mitigation in strongly shaped plasmas and coupling of coaxial helicity injection plasmas to ohmic heating ramp-up. These results advance the ST towards next step fusion energy devices such as NHTX and ST-CTF.

Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

Abstract: Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15–25% n e decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, n e / n GW ∼ 1), to enable n e scan capability (×2) in the H-mode, to test the ability to operate at significantly lower density (e.g., n e / n GW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m 2 ) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

Thermal control of the liquid lithium divertor for NSTX

Abstract: The liquid lithium divertor (LLD) to be installed in NSTX has four toroidal panels, each a conical section inclined at 22° like the previous graphite divertor tiles. Each LLD panel is a copper plate clad with ∼0.25 mm of stainless steel (SS) and a surface layer of flame sprayed molybdenum (Mo) that will host lithium deposited from an evaporator. LITER (evaporators) already used in NSTX will be upgraded for the LLD. Each has twelve 500 W cartridge heaters with thermocouples, 16 other thermocouples, and a channel for helium cooling. During LLD experiments, the LLD will be heated so that the lithium is just above its melting temperature. The length of each shot will be preset to prevent excessive evaporation of lithium from the LLD. This duration depends on the heat load and is likely to be in the range of less than a second to several seconds. Careful thermal control of the LLD is important to maximize the shot times and to guide operation of the LLD. This paper describes the layout of the LLD, its expected thermal performance, the control system, and supporting experiments and analysis. A companion paper in this conference, “Physics design requirements for the national spherical torus experiment liquid lithium divertor,” provides other information.

Mechanical design of the NSTX Liquid Lithium Divertor

Abstract: The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux.