WEST regular in-vessel Inspections with the Articulated Inspection Arm robot

Bruno, V.; Allegretti, L.; Barbuti, A.; Brémond, S.; Cheng, Yong; Dechelle, C.; Dufour, N.; Hao, Feng; Gargiulo, L.; Mouyon, D.; Santraine, B.; Shi, Shanshuang; Song, Yuntao; Sun, Yongjun; Villedieu, E.; Vincent, Benoit; Zheng, Li

doi:10.1016/j.fusengdes.2018.11.050

Cited by 10 publications

(12 citation statements)

References 4 publications

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“…Using the Articulated Inspection Arm [9], a robot able to enter inside WEST without losing vacuum and conditioning, no surface defects have been observed after C4a (10GJ heating power during 9968s, 1157 pulses), although disruptions accompanied by runaway beam have still occurred. This demonstrates the superiority of the bulk W material over the W coatings regarding those very brief and intense events.…”

Section: Discussionmentioning

confidence: 99%

Design and manufacturing of bulk tungsten tiles for WEST outer limiter

Firdaouss

Chappuis

Gazzotti

et al. 2020

Fusion Engineering and Design

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Section: Discussionmentioning

confidence: 99%

Design and manufacturing of bulk tungsten tiles for WEST outer limiter

Firdaouss

Chappuis

Gazzotti

et al. 2020

Fusion Engineering and Design

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“…The campaign was targeted to reach parameters allowing for W fuzz formation in the max OSP area, namely as described in [3] :  an incident He energy above 20 eV (but lower than ~350 eV to avoid competition with erosion, as illustrated in figure 8 of [3])  PFU surface temperature above 900 K (and below 1900 K)  He fluence above the seed fluence required for W fuzz formation (> 10 24 He/m 2 ) A He phase of ~1 week was carried out at the end of the C4 campaign, to allow for PFU post mortem analysis immediately after He exposure.…”

Section: Main Objectives Of the First He Campaign In Westmentioning

confidence: 99%

“…In-vessel inspections using the Articulated Inspection Arm (AIA) of WEST, which allows internal components monitoring during campaigns without breaking the vacuum [24], were performed before and after the He campaign. The first inspection was carried out after the first day dedicated to the changeover from D to He and scenario optimization (12/11/2019) and the second one after the completion of the He campaign (18/11/2019).…”

Section: In Situ Observationsmentioning

confidence: 99%

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Investigation of plasma wall interactions between tungsten plasma facing components and helium plasmas in the WEST tokamak

Tsitrone¹,

Pégouriè²,

Gunn³

et al. 2022

Nucl. Fusion

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ITER will operate with a tungsten divertor, a material featuring surface morphology changes when exposed to helium plasmas, in particular the formation of the so called tungsten fuzz under specific conditions. Investigating interactions between tungsten plasma facing components and helium plasmas in a tokamak environment is therefore a key point to consolidate predictions for the ITER divertor performance and lifetime. To this end, a dedicated helium campaign was performed in the full tungsten WEST tokamak, cumulating ∼2000 s of repetitive L mode discharges. It is shown that conditions for tungsten fuzz formation, as derived from linear devices experiments (incident helium energy E inc > 20 eV, helium fluence >1024 He/m2, surface temperature T surf > 700 °C), were met in the outer strike point (OSP) area of the inertially cooled tungsten divertor elements of WEST. Preliminary inspection of the components after the campaign did not show visible signs of surface modification, but points to significant net erosion in the OSP area. An extensive post mortem analysis is now ongoing to confirm these first indications. These results underline that the complex balance between erosion/redeposition (in particular linked to impurities) and tungsten fuzz formation needs to be taken into account in tokamak conditions.

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“…As metallic plasma-facing materials can be damaged by excessive or uncontrolled heat load, the components lifetime may be reduced and may even affect the plasma operation. The evolution of the WEST PFCs was therefore carefully monitored during and after each campaign through the use of a robotic articulated inspection arm [4] and by carrying out in situ visual inspections during which a portable microscope was used. A particular focus was given to monitoring the ageing of W coatings as local coating failures can lead to the generation of W particles into the plasma and radiative collapse.…”

Section: Introductionmentioning

confidence: 99%

In situ observation of tungsten plasma-facing components after the first phase of operation of the WEST tokamak

et al. 2021

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After three years of operation (2017–2019) and four experimental campaigns (C1 to C4), the first phase of operation of the WEST tokamak has now come to an end. During that period, WEST has operated with tungsten coated copper-based and carbon-based plasma facing components in order to comply with a full metallic environment. A set of components, called plasma facing units, using ITER tungsten monoblock technology was also installed onto one sector of the lower divertor to test their behaviour in a real tokamak environment. This paper gives an overview of the ageing behaviour of these components during in situ visual inspections. It is shown that tungsten coatings are in a good condition after 3 years of operation but are prone to delamination in the areas dominated by off-normal events such as runaway electrons or arcing. For this reason, the W coated carbon tiles of the outer limiter were replaced by bulk W tiles after the C3 campaign. As for the bulk tungsten ITER-like components, damage such as melting and cracking was mainly observed on the monoblock edges exposed to the parallel heat flux. The first observations show that the damage formed during C3 on the tungsten monoblocks did not evolve during C4, which leads to believe that it results from singular events rather than continuous heating due to steady state operation.

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WEST regular in-vessel Inspections with the Articulated Inspection Arm robot

Cited by 10 publications

References 4 publications

Design and manufacturing of bulk tungsten tiles for WEST outer limiter

Design and manufacturing of bulk tungsten tiles for WEST outer limiter

Investigation of plasma wall interactions between tungsten plasma facing components and helium plasmas in the WEST tokamak

In situ observation of tungsten plasma-facing components after the first phase of operation of the WEST tokamak

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