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Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas

文献类型：期刊论文

作者Horacek, J.1; Pitts, R. A.2; Adamek, J.1; Arnoux, G.3; Bak, J-G4; Brezinsek, S.5; Dimitrova, M.1; Goldston, R. J.6; Gunn, J. P.7; Havlicek, J.1,8
刊名PLASMA PHYSICS AND CONTROLLED FUSION
出版日期2016-07-01
卷号58期号:7页码:074005
关键词Tokamak Iter Sol Decay Length Sol Width Scaling
DOI10.1088/0741-3335/58/7/074005
文献子类Article
英文摘要As in many of today's tokamaks, plasma start-up in ITER will be performed in limiter configuration on either the inner or outer midplane first wall (FW). The massive, beryllium armored ITER FW panels are toroidally shaped to protect panel-to-panel misalignments, increasing the deposited power flux density compared with a purely cylindrical surface. The chosen shaping should thus be optimized for a given radial profile of parallel heat flux, q(parallel to) in the scrape-off layer (SOL) to ensure optimal power spreading. For plasmas limited on the outer wall in tokamaks, this profile is commonly observed to decay exponentially as q(parallel to) = q(0)exp (-r/lambda(omp)(q)), or, for inner wall limiter plasmas with the double exponential decay comprising a sharp near-SOL feature and a broader main SOL width, lambda(omp)(q). The initial choice of lambda(omp)(q), which is critical in ensuring that current ramp-up or down will be possible as planned in the ITER scenario design, was made on the basis of an extremely restricted L-mode divertor dataset, using infra-red thermography measurements on the outer divertor target to extrapolate to a heat flux width at the main plasma midplane. This unsatisfactory situation has now been significantly improved by a dedicated multi-machine ohmic and L-mode limiter plasma study, conducted under the auspices of the International Tokamak Physics Activity, involving 11 tokamaks covering a wide parameter range with R = 0.4-2.8 m, B-0 = 1.2-7.5T, I-p = 9-2500 kA. Measurements of lambda(omp)(q) in the database are made exclusively on all devices using a variety of fast reciprocating Langmuir probes entering the plasma at a variety of poloidal locations, but with the majority being on the low field side. Statistical analysis of the database reveals nine reasonable engineering and dimensionless scalings. All yield, however, similar predicted values of lambda(omp)(q) mapped to the outside midplane. The engineering scaling with the highest statistical significance, lambda(omp)(q) = 10(P-tot/V(W m(-3)))(-0.38)(a/R/kappa)(1.3), dependent on input power density, aspect ratio and elongation, yields lambda(omp)(q) = [7, 4, 5] cm for I-p = [2.5, 5.0, 7.5] MA, the three reference limiter plasma currents specified in the ITER heat and nuclear load specifications. Mapped to the inboard midplane, the worst case (7.5 MA) corresponds to lambda(omp)(q) similar to 57 +/- 14 imp mm, thus consolidating the 50 mm width used to optimize the FW panel toroidal shape.
WOS关键词SCRAPE-OFF-LAYER ; TORE-SUPRA TOKAMAK ; TRANSPORT ; TEMPERATURE ; POWER
WOS研究方向Physics
语种英语
WOS记录号WOS:000378616300005
资助机构Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; Czech Science Foundation(GA CR P205/12/2327 ; US DOE(DE-FG02-07ER54917 ; US DOE(DE-FG02-07ER54917 ; US DOE(DE-FG02-07ER54917 ; US DOE(DE-FG02-07ER54917 ; US DOE(DE-FG02-07ER54917 ; US DOE(DE-FG02-07ER54917 ; US DOE(DE-FG02-07ER54917 ; US DOE(DE-FG02-07ER54917 ; US DOE(DE-FG02-07ER54917 ; US DOE(DE-FG02-07ER54917 ; 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源URL[http://ir.hfcas.ac.cn:8080/handle/334002/21836]  
专题合肥物质科学研究院_中科院等离子体物理研究所
作者单位1.Acad Sci Czech Republic, Inst Plasma Phys, Za Slovankou 3, Prague 18000, Czech Republic
2.ITER Org, CS 90 046, F-13067 St Paul Les Durance, France
3.JET, Culham Sci Ctr, EUROfus Consortium, Abingdon OX14 3DB, Oxon, England
4.Natl Fus Res Inst, 113 Yuseong Gu, Daejeon 305333, South Korea
5.Forschungszentrum Julich, D-52425 Julich, Germany
6.Princeton Univ, Plasma Phys Lab, POB 451, Princeton, NJ 08543 USA
7.CEA, IRFM, F-13108 St Paul Les Durance, France
8.Charles Univ Prague, Fac Math & Phys, Prague, Czech Republic
9.MIT, Plasma Sci & Fus Ctr, 175 Albany St, Cambridge, MA 02139 USA
10.Max Planck Inst Plasma Phys, Teilinst Greifswald, D-17491 Greifswald, Germany
推荐引用方式
GB/T 7714		 Horacek, J.,Pitts, R. A.,Adamek, J.,et al. Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas[J]. PLASMA PHYSICS AND CONTROLLED FUSION,2016,58(7):074005.
APA Horacek, J..,Pitts, R. A..,Adamek, J..,Arnoux, G..,Bak, J-G.,...&JET Contributors.(2016).Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas.PLASMA PHYSICS AND CONTROLLED FUSION,58(7),074005.
MLA Horacek, J.,et al."Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas".PLASMA PHYSICS AND CONTROLLED FUSION 58.7(2016):074005.

入库方式： OAI收割

来源：合肥物质科学研究院

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