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NSTX plasma response to lithium coated divertor

Journal of Nuclear Materials

Kugel, H.W.; Bell, M.G.; Allain, J.P.; Bell, R.E.; Ding, S.; Gerhardt, S.P.; Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.M.; Leblanc, B.P.; Maingi, R.; Majeski, R.; Maqueda, R.; Mansfield, D.K.; Mueller, D.; Nygren, Richard E.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Sabbagh, S.A.; Schneider, H.; Skinner, C.H.; Soukhanovskii, V.A.; Taylor, C.N.; Timberlake, J.R.; Wampler, W.R.; Zakharov, L.E.; Zweben, S.J.

NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma-facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, <0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed. © 2010 Elsevier B.V. All rights reserved.

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TYPE Conference YEAR 2011
ScopusOSTI

High heat flux testing of a helium-cooled tungsten tube with porous foam

Fusion Engineering and Design

Youchison, Dennis L.; Lutz, Thomas J.; Williams, B.; Nygren, Richard E.

Utramet, Inc. fabricated one-piece heat exchanger tubes of chemical vapor deposited (CVD) tungsten (W), each with an internal porous mesh fused along either 51 or 38 mm of the axial length of a tube 15 mm in outer diameter. The open porous mesh has a structure of joined ligaments that combines relatively low resistance to flow and a large area for heat transfer. In tests at the Electron Beam Test Stand (EBTS) at Sandia National Laboratories, the maximum absorbed heat load was 22.4 MW/m2 with helium at 4 MPa, flowing at 27 g/s and with inlet and outlet temperatures of 40 and 91 °C and a pressure drop of ∼0.07 MPa. The preparation and testing of the samples was funded through a Phase I grant by the US Department of Energy's Small Business Innovation Research Program. The paper reports the surface temperature distribution indicated by an infrared camera, test conditions, a post-test examination in a scanning electron microscope and other details. © 2007 Elsevier B.V. All rights reserved.

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TYPE Conference YEAR 2007
ScopusOSTI

Features of plasma sprayed beryllium armor for the ITER first wall

Journal of Nuclear Materials

Nygren, Richard E.; Youchison, Dennis L.; Hollis, K.J.

Two water-cooled mockups with CuCrZr heat sinks and plasma sprayed beryllium (PS Be) armor, 5 and 10 mm thick respectively, were fabricated at Los Alamos National Laboratory and thermally cycled at Sandia at 1 and 2 MW/m2. The castellated surface of the CuCrZr mechanically locked the armor. The resulting PS Be morphology controlled cracking during thermal cycling. Post test examinations showed transverse cracks perpendicular to the surface of the armor that would relieve thermal stresses but not degrade heat transfer. The mockups and two others previously produced for the European Fusion Development Agreement had somewhat porous armor, with a thermal conductivity estimated to be about 1/4 that of fully dense beryllium, due to the low (600-650 °C) substrate temperature during deposition specifically requested by EFDA to avoid subsequent heat treating of CuCrZr. Some melting of the armor was expected and observed in the tests. © 2007 Elsevier B.V. All rights reserved.

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TYPE Conference YEAR 2007
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Results 51–75 of 89
Results 51–75 of 89