In ITER, as in any tokamak, the first wall and divertor plasma-facing components (PFC) must provide adequate protection of in-vessel structures, sufficient heat exhaust capability and be compatible ...with the requirements of plasma purity. These functions take on new significance in ITER, which will combine long pulse, high power operation with severe restrictions on permitted core impurity concentrations and which, in addition, will produce transient energy loads on a scale unattainable in today’s devices. The current ITER PFC design has now reached a rather mature stage following the 2007 ITER Design Review. This paper presents the key elements of the design, reviews the physics drivers, essentially thermal load specifications, which have defined the concept and discusses a selection of material and design issues.
Budget restrictions have forced the ITER Organization to reconsider the baseline divertor strategy, in which operations would begin with carbon (C) in the high heat flux regions, changing out to a ...full-tungsten (W) variant before the first nuclear campaigns. Substantial cost reductions can be achieved if one of these two divertors is eliminated. The new strategy implies not only that ITER would start-up on a full-W divertor, but that this component should survive until well into the nuclear phase. This paper considers the risks engendered by such an approach with regard to known W plasma-material interaction issues and briefly presents the current status of a possible full-W divertor design.
•Detailed design development plan for the ITER tungsten divertor.•Latest status of the ITER tungsten divertor design.•Brief overview of qualification program for the ITER tungsten divertor and status ...of R&D activity.
In November 2011, the ITER Council has endorsed the recommendation that a period of up to 2 years be set to develop a full-tungsten divertor design and accelerate technology qualification in view of a possible decision to start operation with a divertor having a full-tungsten plasma-facing surface. To ensure a solid foundation for such a decision, a full tungsten divertor design, together with a demonstration of the necessary high performance tungsten monoblock technology should be completed within the required timescale. The status of both the design and technology R&D activity is summarized in this paper.
•Infrared quantitative thermography challenging in fully reflective and radiative environment.•synthetic diagnostic is a remarkable tool to reduce the risk of misleading interpretation of IR ...image.•dedicated experiences to measure the optical properties of fusion materials.•comparison of simulated and experimental data in WEST and ASDEX-U devices.
Infra-red (IR) thermography is a widely used tool in fusion devices to monitor and to protect the plasma-facing component (PFC) from excessive heat loads. However, with the use of all-metal walls in fusion devices, deriving surface temperature from IR measurements has become more challenging. In this paper, an overview of infra-red measurements in the metallic tokamaks WEST and ASDEX Upgrade (AUG) is reported and the techniques carried out in the modeling and experimental fields to deal with this radiative and fully reflective environment are presented. Experimental characterizations of metallic samples in laboratory and experiments in WEST and AUG reveal that the behavior of both the emission and the reflectance can vary significantly with surface roughness, machining process and as the plasma operation progress. In parallel, the development of a synthetic IR diagnostic has allowed for a better interpretation of the IR images by assessing the reflection patterns and their origin. This has also proven that small-scale change in the emission pattern of beveled PFC can be confused with abnormal thermal events. Numerical solutions to evaluate the contribution of the reflections associated with a variable emissivity in a fully reflective and radiative environment are finally presented.
A concept for a shaped first wall for ITER is presented. While keeping most features of the 2004 FDR wall (modules segmentation, plasma facing components technologies, plasma facing material), this ...concept provides protection of the lateral faces of the first wall panels against the intense parallel heat flux coming from the plasma. Excessive beryllium temperatures at the panel edges are avoided during regular operation. The intense heat flux at the top of the vessel is accounted for and protection is provided against the shine thru heat flux. Start-up and ramp down using the wall as a limiter is possible for up to 7.5
MW, both inboard and outboard. This is rendered possible by the use of 5
MW/m
2 technology panels for 40% of the panels.
The main key to achieving high-power long-duration discharges on Tore Supra, the actively cooled toroidal pump limiter (TPL) is the main plasma-facing component, handling high heat fluxes. The heat ...pattern on the TPL presents features of both localized and large-area limiters (mixed influences of parallel and cross-field heat fluxes). The combination of the toroidal field ripple and the flat surface results in a peaked heat flux pattern with large private flux areas on the surface. The apparent heat flux decay length is shorter than 10 mm and varies by less than 10% with the plasma conditions. The conduction/convection is modeled within 5% by the heat flux deposition code TOKAFLUX. The heat pattern is further modified by the contribution of suprathermal particles (ion ripple losses, fast electrons). Altogether, the relation of the peak heat flux to a given injected power is consistent with modeling made during TPL design. The thermal response of the elements is also in line with the design, with a typical thermal time constant of 1 s and steady-state surface temperature during long discharges. An important issue being investigated concerns the growth of material deposits; they accumulate in shadowed areas and especially just along the frontier to plasma-wetted areas. In 2009, the limiter is still in operation and several thematics are still being actively investigated, such as the effect of the material deposits on the operation, the long-time-scale behavior of the tile to heat sink bond, and the deuterium retention.
ITER plasma-facing components Merola, Mario; Loesser, D.; Martin, A. ...
Fusion engineering and design,
12/2010, Letnik:
85, Številka:
10
Journal Article, Conference Proceeding
Recenzirano
The ITER plasma-facing components directly face the thermonuclear plasma and include the divertor, the blanket and the test blanket modules with their corresponding frames.
The divertor is located at ...the bottom of the plasma chamber and is aimed at exhausting the major part of the plasma thermal power (including alpha power) and at minimising the helium and impurity content in the plasma.
The blanket system provides a physical boundary for the plasma transients and contributes to the thermal and nuclear shielding of the vacuum vessel and external machine components. It consists of modular shielding elements known as blanket modules which are attached to the vacuum vessel. Each blanket module consists of two major components: a plasma-facing first wall panel and a shield block.
The test blanket modules are mock-ups of DEMO breeding blankets. There are three ITER equatorial ports devoted to test blanket modules, each of them providing for the allocation of two breeding modules inserted in a steel frame and in front of a shield block.
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