Full coverage infrared thermography diagnostic for WEST machine protection

Courtois, X.; Aumeunier, M.-H.; Balorin, C.; Migozzi, J. B.; Houry, M.; Blanckaert, Koen; Moudden, Y.; Pocheau, C.; Saille, A.; Hugot, Emmanuel; Marcos, Michel; Vivès, S.

doi:10.1016/j.fusengdes.2019.03.090

Cited by 33 publications

(23 citation statements)

References 11 publications

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“…The W-coated graphite is divided in two separate components, one in the low field side (LFS/outer) and one in the high field side (HFS/inner). The WEST lower divertor is monitored by a set of complementary diagnostics allowed characterizing properties of heat loads on the lower divertor: 20 embedded TCs located at 7.5mm below the surface [10], 4 FBGs [3] embedded at 3.5 and 7mm below the surface, IR thermography [4] and two arrays of 29 flush-mounted LPs [5].The diagnostic are spread over the 12 divertor sectors keeping symmetric toroidal location of the sensors regarding the magnetic field toroidal modulation (ripple effect). A poloidal cross section of the WEST device with two magnetic configurations with a separatix located at Rext = 2.93m and 3m is shown in fig 1 a.…”

Section: Heat Load Diagnostics Set-up and Heat Flux Databasementioning

confidence: 99%

Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST

et al. 2021

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WEST is a full W tokamak with an extensive set of diagnostics for heat load measurements especially in the lower divertor. It is composed by infrared (IR) thermography, thermal measurement with thermocouples (TC) and fiber Bragg grating (FBG) embedded few mm below the surface and flush mounted Langmuir probes (LP). A large database including different magnetic equilibrium and input power is investigated to compare the heat load pattern (location, amplitude of the peak and heat flux decay length) on the inner and outer strike point regions as function of the continuous progress achieved in WEST : from the first ohmic diverted plasma (obtained during the second experimental campaign C2 in 2018) up to the high power (up to 8 MW total injected) and high energy (up to 90 MJ injected energy in lower single null configuration) steady state experiments performed in the last experimental campaign (C4 in 2019). Concerning the peak location, a good agreement (<1cm) is obtained between thermal inversions and flush-mounted LP measurements. The peak heat flux from the whole set of diagnostics is in good agreement and mainly in the ± 20% range, while the heat flux decay length reported on the target shows significant discrepancy between diagnostics and location in the machine (± 40% range). Despite such discrepancy, heat flux decay length at target is found to scale mainly with the magnetic flux expansion through the variation of the X-point height, as expected. The improved plasma performances achieved during C4 enabled to reach significant heat load in the divertor, up to 6 MW/m² with 4 MW of additional heating power showing the capability to reach the ITER relevant heat load (10 MW/m² steady state) with about 7 MW of additional power in L-mode discharge. The heat load distribution is clearly asymmetric with a 3/4 and 1/4 distribution on the outer and inner strike point region respectively for the parallel heat flux close to the 2/3 -1/3 commonly observed in forward-B divertor configuration.

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Section: Heat Load Diagnostics Set-up and Heat Flux Databasementioning

confidence: 99%

Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST

et al. 2021

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“…How these impurity seeding experiments might scale to the SPARC device and its effect on the ability to achieve the primary mission goal of Q > 2 is unclear because of the sizable extrapolation to SPARC divertor parameters from these devices. Alternatively, active feedback control systems have also been developed that can reduce the heat flux to the divertor target surface without negatively affecting core plasma performance (Kallenbach et al 2012;Brunner et al 2017;Guillemaut et al 2017;Courtois et al 2019). Whether any of these active-feedback systems can be implemented in the SPARC device has yet to be assessed, but it will be challenging given the available diagnostic access in the nuclear environment and the actuator response times.…”

Section: Introductionmentioning

confidence: 99%

“…2017; Courtois et al . 2019). Whether any of these active-feedback systems can be implemented in the SPARC device has yet to be assessed, but it will be challenging given the available diagnostic access in the nuclear environment and the actuator response times.…”

Section: Introductionmentioning

confidence: 99%

Divertor heat flux challenge and mitigation in SPARC

et al. 2020

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Owing to its high magnetic field, high power, and compact size, the SPARC experiment will operate with divertor conditions at or above those expected in reactor-class tokamaks. Power exhaust at this scale remains one of the key challenges for practical fusion energy. Based on empirical scalings, the peak unmitigated divertor parallel heat flux is projected to be greater than 10 GW m−2. This is nearly an order of magnitude higher than has been demonstrated to date. Furthermore, the divertor parallel Edge-Localized Mode (ELM) energy fluence projections (~11–34 MJ m−2) are comparable with those for ITER. However, the relatively short pulse length (~25 s pulse, with a ~10 s flat top) provides the opportunity to consider mitigation schemes unsuited to long-pulse devices including ITER and reactors. The baseline scenario for SPARC employs a ~1 Hz strike point sweep to spread the heat flux over a large divertor target surface area to keep tile surface temperatures within tolerable levels without the use of active divertor cooling systems. In addition, SPARC operation presents a unique opportunity to study divertor heat exhaust mitigation at reactor-level plasma densities and power fluxes. Not only will SPARC test the limits of current experimental scalings and serve for benchmarking theoretical models in reactor regimes, it is also being designed to enable the assessment of long-legged and X-point target advanced divertor magnetic configurations. Experimental results from SPARC will be crucial to reducing risk for a fusion pilot plant divertor design.

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“…An infrared thermography system monitors the antenna front-faces with a spatial resolution better than 5 mm/pixel [15]. Antenna limiters are protected with a real-time reduction of the RF power if the luminance temperature (i.e.…”

Section: Introduction and Antennas Descriptionmentioning

confidence: 99%

WEST actively cooled load resilient ion cyclotron resonance heating system results

Hillairet¹,

Mollard²,

Colas³

et al. 2021

Nucl. Fusion

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Three identical new WEST Ion Cyclotron Resonance Heating (ICRH) antennas have been designed, assembled then commissioned on plasma from 2013 to 2019. The WEST ICRH system is both load-resilient and compatible with long-pulse operations. The three antennas have been successfully operated together on plasma in 2019 and 2020. The load resilience capability has been demonstrated and the antenna feedback controls for phase and matching have been developed. The breakdown detection systems have been validated and successfully protected the antennas. The use of ICRH in combination with Lower Hybrid has triggered the first high confinement mode transitions identified on WEST.

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Full coverage infrared thermography diagnostic for WEST machine protection

Cited by 33 publications

References 11 publications

Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST

Divertor power loads and scrape off layer width in the large aspect ratio full tungsten tokamak WEST

Divertor heat flux challenge and mitigation in SPARC

WEST actively cooled load resilient ion cyclotron resonance heating system results

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