Abstract
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, 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 , or, for inner wall limiter plasmas with the double exponential decay comprising a sharp near-SOL feature and a broader main SOL width, . The initial choice of , 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 Measurements of 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 mapped to the outside midplane. The engineering scaling with the highest statistical significance, , dependent on input power density, aspect ratio and elongation, yields = [7, 4, 5] cm for = [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 mm, thus consolidating the 50 mm width used to optimize the FW panel toroidal shape.
Highlights
As in many tokamaks, ITER will use the main chamber first wall (FW) as a limiter for a part of the plasma ramp-up and down phases
This paper presents the results of this database activity, develops a variety of possible scalings from the data and provides recommendations for the values of λqomp to be used in the ITER FW shaping design
To hide misalignments between neighbouring plasma-facing panels on the ITER main chamber FW, the panels are toroidally shaped. This shaping increases the heat flux on the panel compared with an ideal cylindrical surface and this is a key disadvantage given the use of beryllium as armour material and the use of active cooling, which limit the steady state power handling capacity for limited plasmas which will characterize the start-up and ramp-down phases of all ITER plasmas
Summary
ITER will use the main chamber first wall (FW) as a limiter for a part of the plasma ramp-up and down phases. The ITER FW consists of 440 blanket modules, each comprising a massive steel shield block protected by a beryllium armoured panel The latter is constituted of poloidal arrays of toroidally extended, water cooled fingers made up of Be flat tiles bonded to a copper chromium zirconium heat sink, itself bonded to a steel structure bearing the water cooling channels [2]—see figures 2(a) and (b). Central to the shape design is the value of λqomp, which is not known for ITER and which must be chosen either from a scaling or physics-based appproach It must, be a fixed value with sufficient margin to encompass the range of plasma currents and magnetic configurations envisaged for limiter plasmas on ITER (see below). J Horacek et al framework based on quasi-linear transport theory which agrees very favourably with a subset of the ITPA database presented in this paper
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