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Title: Divertor Heat Flux Amelioration in Highly-Shaped Plasma in NSTX

Abstract

Steady-state handling of divertor heat flux is a critical issue for both the International Thermonuclear Experimental Reactor and spherical torus (ST) based devices with compact high power density divertors. The ST compact divertor with a small plasma volume, a small plasma-wetted area, and a short parallel connection length can reduce the operating space of heat flux dissipation techniques based on induced edge and/or scrape-off layer (SOL) power and momentum loss, such as the radiative and dissipative divertors and radiative mantles. Access to these regimes is studied in the National Spherical Torus Experiment (NSTX) with an open geometry horizontal carbon plate divertor in 2-6 MW NBI-heated H-mode plasmas in a lower single null (LSN) configuration in a range of elongations {kappa} = 1.8-2.4 and triangularities {delta}= 0.40-0.75. Experiments conducted in a lower end {kappa}{approx}1.8-2.0 and {delta}{approx} 0.4-0.5 LSN shape using deuterium injection in the divertor region have achieved the outer strike point (OSP) peak heat flux reduction from 4-6 MW/m2 to a manageable level of 1-2 MW/m2. However, only the high-recycling radiative divertor (RD) regime was found to be compatible with good performance and H-mode confinement. A partially detached divertor (PDD) could only be obtained at a high D2 injection ratemore » that led to an X-point MARFE formation and confinement degradation. Also in the low {kappa}{approx} 2,{delta}{approx} 0.45 shape, peak heat flux q{sub pk} and heat flux width {lambda}{sub q} scaling studies have been conducted. Similar to tokamak divertor studies, q{sub pk} was found to be a strong function of input power PNBI and plasma current Ip, and the heat flux midplane scale length {lambda}{sub q} was found to be large as compared with simple SOL models. In this paper, we report on the first experiments to assess steady-state divertor heat flux amelioration in highly shaped plasmas in NSTX.« less

Authors:
; ; ; ; ; ; ; ; ; ; ;
Publication Date:
Research Org.:
Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
957600
Report Number(s):
UCRL-CONF-232417
TRN: US1005578
DOE Contract Number:  
W-7405-ENG-48
Resource Type:
Conference
Resource Relation:
Conference: Presented at: 34th EPS Conference on Plasma Physics, Warsaw, Poland, Jul 02 - Jul 06, 2007
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION; CARBON; CONFIGURATION; CONFINEMENT; DEUTERIUM; DIVERTORS; ELECTRIC CURRENTS; EXPERIMENTAL REACTORS; GEOMETRY; HEAT FLUX; MARFE; PERFORMANCE; PHYSICS; PLASMA; PLATES; POWER DENSITY; SHAPE; SOLS

Citation Formats

Soukhanovskii, V, Maingi, R, Gates, D, Menard, J, Raman, R, Bell, R, Bush, C, Kaita, R, Kugel, H, LeBlanc, B, Paul, S, and Roquemore, A. Divertor Heat Flux Amelioration in Highly-Shaped Plasma in NSTX. United States: N. p., 2007. Web.
Soukhanovskii, V, Maingi, R, Gates, D, Menard, J, Raman, R, Bell, R, Bush, C, Kaita, R, Kugel, H, LeBlanc, B, Paul, S, & Roquemore, A. Divertor Heat Flux Amelioration in Highly-Shaped Plasma in NSTX. United States.
Soukhanovskii, V, Maingi, R, Gates, D, Menard, J, Raman, R, Bell, R, Bush, C, Kaita, R, Kugel, H, LeBlanc, B, Paul, S, and Roquemore, A. Mon . "Divertor Heat Flux Amelioration in Highly-Shaped Plasma in NSTX". United States. https://www.osti.gov/servlets/purl/957600.
@article{osti_957600,
title = {Divertor Heat Flux Amelioration in Highly-Shaped Plasma in NSTX},
author = {Soukhanovskii, V and Maingi, R and Gates, D and Menard, J and Raman, R and Bell, R and Bush, C and Kaita, R and Kugel, H and LeBlanc, B and Paul, S and Roquemore, A},
abstractNote = {Steady-state handling of divertor heat flux is a critical issue for both the International Thermonuclear Experimental Reactor and spherical torus (ST) based devices with compact high power density divertors. The ST compact divertor with a small plasma volume, a small plasma-wetted area, and a short parallel connection length can reduce the operating space of heat flux dissipation techniques based on induced edge and/or scrape-off layer (SOL) power and momentum loss, such as the radiative and dissipative divertors and radiative mantles. Access to these regimes is studied in the National Spherical Torus Experiment (NSTX) with an open geometry horizontal carbon plate divertor in 2-6 MW NBI-heated H-mode plasmas in a lower single null (LSN) configuration in a range of elongations {kappa} = 1.8-2.4 and triangularities {delta}= 0.40-0.75. Experiments conducted in a lower end {kappa}{approx}1.8-2.0 and {delta}{approx} 0.4-0.5 LSN shape using deuterium injection in the divertor region have achieved the outer strike point (OSP) peak heat flux reduction from 4-6 MW/m2 to a manageable level of 1-2 MW/m2. However, only the high-recycling radiative divertor (RD) regime was found to be compatible with good performance and H-mode confinement. A partially detached divertor (PDD) could only be obtained at a high D2 injection rate that led to an X-point MARFE formation and confinement degradation. Also in the low {kappa}{approx} 2,{delta}{approx} 0.45 shape, peak heat flux q{sub pk} and heat flux width {lambda}{sub q} scaling studies have been conducted. Similar to tokamak divertor studies, q{sub pk} was found to be a strong function of input power PNBI and plasma current Ip, and the heat flux midplane scale length {lambda}{sub q} was found to be large as compared with simple SOL models. In this paper, we report on the first experiments to assess steady-state divertor heat flux amelioration in highly shaped plasmas in NSTX.},
doi = {},
url = {https://www.osti.gov/biblio/957600}, journal = {},
number = ,
volume = ,
place = {United States},
year = {2007},
month = {7}
}

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