Status of fabrication development for plasma facing components in the EU

Daenner, W.; Merola, M.; Lorenzetto, P.; Peacock, Alan T.; Bobin-Vastra, I.; Briottet, Laurent; Bucci, P.; Conchon, D; Erskine, A; Escourbiac, F.; Febvre, M; Grattarola, M.; Hjorth, Carl Gustaf; Hofmann, G.; Ilzhoefer, A.; Lill, K; Lind, A.; Linke, J.; Richards, Will; Rigal, Emmanuel; Roedig, M.; Saint-Antonin, François; Schedler, B.; Schlosser, J.; Tähtinen, Seppo; Visca, E.

doi:10.1016/s0920-3796(02)00159-x

Cited by 26 publications

(11 citation statements)

References 7 publications

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“…23,24) CFC materials have been developed with the goal to obtain high thermal conductivity (>300 W/mK at RT), high density, high purity and a proper balance of the mechanical properties. Details concerning the development of CFC for fusion applications are described in Ref.…”

Section: Cfc Materials and The Componentsmentioning

confidence: 99%

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

2005

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The current ITER design employs beryllium, carbon fiber reinforced composite and tungsten as plasma facing materials. Since these materials are exposed to high heat fluxes during the operation, it is essential to perform high heat flux tests for R&D of ITER components. Static heat loads corresponding to cycling loads during normal operation, are estimated to be up to 20 MW/m 2 in the divertor targets and around 0.5 MW/m 2 at the first wall in ITER. For the static high heat flux testing, tests in electron beam facilities, particle beam facilities, IR heater and in-pile tests have been performed. Another type, more critical heat loads, which have high power densities and short durations, corresponding to transient events, i.e. plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux onto the plasma facing materials. For this purpose, tests in electron beam (short pulses), plasma gun and high power laser facilities have been carried out. The present work summarizes the features of these facilities and recent experimental results as well as the current selection of ITER plasma facing components.

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Section: Cfc Materials and The Componentsmentioning

confidence: 99%

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

2005

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“…To alleviate the thermo-mechanical constraints and to ensure the durability of the PFCs it was decided to split the first wall and divertor into a large number of compact cells -by introducing the so-called castellated structures [1,2]. However, along with the favorable significant reduction of the risk of the PFC failure, the use of castellation has triggered a number of critical issues:…”

Section: Introductionmentioning

confidence: 99%

Overview of material migration and mixing, fuel retention and cleaning of ITER-like castellated structures in TEXTOR

Litnovsky

Kreter

Kirschner

et al. 2011

Journal of Nuclear Materials

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Plasma-facing components (PFCs) in ITER will be castellated by splitting them into small-size blocks to maintain the thermo-mechanical stability. However, there are concerns in 2 particular on retention of codeposited radioactive fuel in the gaps. An R&D program is underway in TEXTOR addressing this acute issue of castellation. Material migration and fuel inventory are investigated using long-and short-term discharge-resolved experiments with castellated structures in TEXTOR. Significant impurity transport to the gaps was detected and results were in part quantitatively reproduced with 3D-GAPS code.

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“…In ITER the armor of the first wall and divertor area will be castellated to maintain the thermomechanical durability of the machine [1,2]. Given the difficulties in fuel removal [3][4][5], there is a concern that the radioactive fuel will be accumulated in the gaps of such a castellation [6][7][8][9].…”

Section: Introductionmentioning

confidence: 99%

Castellated structures for ITER: Differences of impurity deposition and fuel accumulation in the toroidal and poloidal gaps

Litnovsky¹,

Philipps²,

Wienhold

et al. 2009

Journal of Nuclear Materials

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Castellation is foreseen for the first wall and divertor area in ITER. The concern of the fuel accumulation and impurity deposition in the gaps of castellated structures calls for dedicated studies.Recently, a tungsten castellated limiter with rectangular and roof-like shaped cells was exposed to the SOL plasmas in TEXTOR. After exposure, roughly two times less fuel was found in the gaps between the shaped cells whereas the difference in carbon deposition was less pronounced. Up to 70 at. % of tungsten was found intermixed in the deposited layers in the gaps. The metal fraction in the deposit decreases rapidly with a depth of the gap. Modeling of carbon deposition in poloidal gaps has provided a qualitative agreement with an experiment. Misalignment of toroidal gaps with respect to the magnetic field lines has led to the significant anisotropy of C and D distribution in the gaps.

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Status of fabrication development for plasma facing components in the EU

Cited by 26 publications

References 7 publications

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

Overview of material migration and mixing, fuel retention and cleaning of ITER-like castellated structures in TEXTOR

Castellated structures for ITER: Differences of impurity deposition and fuel accumulation in the toroidal and poloidal gaps

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