Status of the ITER full-tungsten divertor shaping and heat load distribution analysis

Carpentier-Chouchana, S.; Hirai, T.; Escourbiac, F.; Durocher, A.; Fedosov, A. M.; Ferrand, L.; Firdaouss, M.; Kočan, M.; Kukushkin, A.S.; Jokinen, Tommi; Komarov, V.; Lehnen, M.; Merola, M.; Mitteau, R.; Pitts, R.A.; Stangeby, P.C.; Sugihara, M.

doi:10.1088/0031-8949/2014/t159/014002

Cited by 49 publications

(27 citation statements)

References 5 publications

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“…In ITER, PFCs will be castellated, i.e. split into small segments separated by thin gaps [21,22]. The gap entrance corresponds to a small fraction of the plasma-exposed area (the gap width is 0.5 mm), implying that dust trajectories directly terminating in the gaps can be considered rare.…”

Section: Introductionmentioning

confidence: 99%

Dust remobilization in fusion plasmas under steady state conditions

Tolias

Ratynskaia

Angeli

et al. 2016

Plasma Phys. Control. Fusion

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The first combined experimental and theoretical studies of dust remobilization by plasma forces are reported. The main theoretical aspects of remobilization in fusion devices under steady state conditions are analyzed. In particular, the dominant role of adhesive forces is highlighted and generic remobilization conditions -direct lift-up, sliding, rolling -are formulated. A novel experimental technique is proposed, based on controlled adhesion of dust grains on tungsten samples combined with detailed mapping of the dust deposition profile prior and post plasma exposure. Proof-ofprinciple experiments in the TEXTOR tokamak and the EXTRAP-T2R reversed-field pinch are presented. The versatile environment of the linear device Pilot-PSI allowed for experiments with different magnetic field topologies and varying plasma conditions that were complemented with camera observations.

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Section: Introductionmentioning

confidence: 99%

Dust remobilization in fusion plasmas under steady state conditions

Tolias

Ratynskaia

Angeli

et al. 2016

Plasma Phys. Control. Fusion

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“…Values of intra-PFU gap width g MB and radial misalignment m rad , and inter-PFU gap width g PFU and radial misalignment m PFU used in this study 1. …”

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confidence: 99%

A study of planar toroidal–poloidal beveling of monoblocks on the ITER divertor outer vertical target

et al. 2019

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The design of the monoblocks constituting the ITER divertor vertical targets comprises a simple toroidal (i.e. toroidally-facing) bevel of 0.5 mm in order to magnetically shadow poloidal (i.e. poloidally-running) leading edges, arising from radial misalignments between toroidally neighbouring blocks, from parallel heat loads between and during ELMs. Previous studies suggest that excessive heating of long toroidal edges could also occur, possibly leading to melting during ELMs. Furthermore, despite the toroidal bevel, tiny regions of the poloidal leading edges known as "optical hot spots", accessible along magnetic field lines through toroidal gaps, remain exposed to parallel heat flux from ELMs. The intense heat flux onto those optical hot spots could be large enough to trigger tungsten boiling. A possible solution at the outer vertical target is to implement a planar toroidal-poloidal bevel that would hide all poloidal and toroidal edges and eliminate the optical hot spot. It will be demonstrated that a reasonable "shallow" toroidal-poloidal bevel solution solves all these problems with minimal trade-offs, under the condition that monoblocks on neighbouring plasma-facing units be well aligned poloidally in order to prevent the appearance of exposed leading edges, meaning, in the worst case, a stepwise downward shift of each toroidally upstream plasmafacing unit by -2±2 mm with respect to their downstream neighbours. A more deeply beveled solution has also been studied that is immune to poloidal misalignments, but which comprises important trade-offs in terms of higher heat load to the main wetted surface, and excessive ELM heat loads onto the magnetically shadowed side of the toroidal gaps. Unfortunately, due the inclination of magnetic flux surfaces, the planar toroidal-poloidal beveling solution does not work at the inner vertical target, meaning that its application at the outer target alone leaves the inner toroidal gaps unprotected. This, together with the technologically challenging requirement for a high degree of poloidal alignment of toroidally neighbouring plasma-facing units, has led to a decision not to apply the poloidal-toroidal bevel solution on the ITER vertical targets.

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“…8 The WEST project consists in transforming the Tore Supra tokamak in an X-point divertor configuration while keeping its long pulse capability, in order to test the ITER divertor technology. 8 The main goal is to test the actively cooled tungsten (W) Plasma Facing Units (PFUs) representative of the ITER divertor target 9,10 under high heat load (in the 10-20 MW m −2 range). In WEST, each PFU is composed of 35 W monoblocks of individual size 28 mm × 12 mm × 26 mm (width × depth × height) assembled with a gap of 0.5 mm on a CuCrZr heat sink tube.…”

Section: Introductionmentioning

confidence: 99%

Integration of fiber Bragg grating temperature sensors in plasma facing components of the WEST tokamak

Corre

Laffont

Pocheau

et al. 2018

Review of Scientific Instruments

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Plasma Facing Components (PFC) temperature measurement is mandatory to ensure safe high power and long pulse tokamak operation. IR thermography systems which are widely used in magnetic fusions devices become challenged with the choice of tungsten as a PFC material in the ITER tokamak, mainly due to emissivity uncertainties and reflection issues in a hot environment. Embedded temperature measurements are foreseen to cross-check the IR thermography measurements. Fiber Bragg grating sensors are potentially of great interest for this application because they are immune to electromagnetic interference and allow the measurement of a large number of temperature spots on a single fiber. Four optical fiber temperature sensing probes, each of them including 11 regenerated fiber Bragg gratings equally spaced by 12.5 mm (equivalent to one ITER-like tungsten monoblock), have been specifically designed and manufactured for the WEST project (W-tungsten Environment and Steady State Tokamak). The four probes are embedded in W-coated graphite components at two different distances from the surface, 3.5 mm and 7 mm, to cover a wide range of temperatures up to 900 °C. This paper addresses the design and integration issues and the qualification and performance assessment performed in the laboratory. It also shows the first measurements of this new diagnostic achieved in a tokamak environment during baking of the machine and during early diverted plasma exposure.

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Status of the ITER full-tungsten divertor shaping and heat load distribution analysis

Cited by 49 publications

References 5 publications

Dust remobilization in fusion plasmas under steady state conditions

Dust remobilization in fusion plasmas under steady state conditions

A study of planar toroidal–poloidal beveling of monoblocks on the ITER divertor outer vertical target

Integration of fiber Bragg grating temperature sensors in plasma facing components of the WEST tokamak

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