Overview of the recent DiMES and MiMES experiments in DIII-D

Rudakov, D.L.; Wong, C.P.C.; Litnovsky, A.; Wampler, W.R.; Boedo, J.A.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Hollmann, E. M.; Jacob, W.; Krasheninnikov, S. I.; Krieger, K.; Lasnier, C.J.; Leonard, A. W.; McLean, A.G.; Marot, L.; Moyer, R. A.; Pétrie, T.W.; Philipps, V.; Smirnov, R.D.; Stangeby, P.C.; Watkins, J.G.; West, W.P.; Yu, J. H.

doi:10.1088/0031-8949/2009/t138/014007

Cited by 20 publications

(12 citation statements)

References 23 publications

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“…In the US, both DIII-D and NSTX, NSTX-U have the possibility to introduce materials into the divertor. For DIII-D the so-called DiMES manipulator [197] is introduced into the flat floor of the divertor [198][199][200] allowing exposure, observation and exchange of material samples. For NSTX and NSTX-U a materials test station (MAPP) can be used to also analyze samples in-vacuo after exposure by means of e.g.…”

Section: Materials Tests In Tokamaksmentioning

confidence: 99%

“…Even though it is yet unclear how much of the material that is eroded eventually converts into dust, studies have focused recently on the dynamics of dust particulates observed in tokamaks and stellarators with a size range between 10 nm and 100 μm [314,315]. Studies have been performed in several tokamaks covering both limiter configurations such as TEXTOR [192,316] and divertor machines like DIII-D [198,317]. Moreover, studies how to detect and study dust in ITER were undertaken [206].…”

Section: Erosion-migration-depositionmentioning

confidence: 99%

“…Moreover, studies how to detect and study dust in ITER were undertaken [206]. In order to allow for controlled particle sizes and easy implementation of experiments, dust studies have been performed utilizing either limiter lock systems [192] or the DiMES manipulator [198].…”

Section: Erosion-migration-depositionmentioning

confidence: 99%

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Material testing facilities and programs for plasma-facing component testing

et al. 2017

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Component development for operation in a large-scale fusion device requires thorough testing and qualification for the intended operational conditions. In particular environments are necessary which are comparable to the real operation conditions, allowing at the same time for in situ/in vacuo diagnostics and flexible operation, even beyond design limits during the testing. Various electron and neutral particle devices provide the capabilities for high heat load tests, suited for material samples and components from lab-scale dimensions up to full-size parts, containing toxic materials like beryllium, and being activated by neutron irradiation. To simulate the conditions specific to a fusion plasma both at the first wall and in the divertor of fusion devices, linear plasma devices allow for a test of erosion and hydrogen isotope recycling behavior under well-defined and controlled conditions. Finally, the complex conditions in a fusion device (including the effects caused by magnetic fields) are exploited for component and material tests by exposing test mock-ups or material samples to a fusion plasma by manipulator systems. They allow for easy exchange of test pieces in a tokamak or stellarator device, without opening the vessel. Such a chain of test devices and qualification procedures is required for the development of plasma-facing components which then can be successfully operated in future fusion power devices. The various available as well as newly planned devices and test stands, together with their specific capabilities, are presented in this manuscript. Results from experimental programs on test facilities illustrate their significance for the qualification of plasma-facing materials and components. An extended set of references provides access to the current status of material and component testing capabilities in the international fusion programs.

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Section: Materials Tests In Tokamaksmentioning

confidence: 99%

Section: Erosion-migration-depositionmentioning

confidence: 99%

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Material testing facilities and programs for plasma-facing component testing

et al. 2017

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“…The RG castellation was mounted in the DiMES head in the similar manner as described in [6]. In total 29 discharges were made with the total plasma duration of $100 s. The castellation was exposed to colder and less dense plasmas: N e $ 1.5 ⁄ 10 12 cm À3 , T e $ 15 eV than that during exposure in TEXTOR.…”

Section: Materials Migration In the Gapsmentioning

confidence: 99%

Optimization of tungsten castellated structures for the ITER divertor

Litnovsky

Hellwig

Matveev

et al. 2015

Journal of Nuclear Materials

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a b s t r a c tIn ITER, the plasma-facing components (PFCs) of the first wall and the divertor armor will be castellated to improve their thermo-mechanical stability and to limit forces due to induced currents. The fuel accumulation in the gaps may significantly contribute to the in-vessel fuel inventory. Castellation shaping may be the most straightforward way to minimize the fuel inventory and to alleviate the thermal loads onto castellations.A new castellation shape was proposed and comparative modeling of conventional (rectangular) and shaped castellation was performed for ITER conditions. Shaped castellation was predicted to be capable to operate under stationary heat load of 20 MW/m 2 . An 11-fold decrease of beryllium (Be) content in the gaps of the shaped cells alone with a 7-fold decrease of carbon content was predicted. In order to validate the predictive capabilities of modeling tools used for ITER conditions, the dedicated modeling with the same codes was made for existing tokamaks and benchmarked with the results of multi-machine experiments. For the castellations exposed in TEXTOR and DIII-D, the carbon amount in the gaps of shaped cells was 1.9-2.3 times smaller than that of rectangular ones. Modeling for TEXTOR conditions yielded to 1.5-fold decrease of carbon content in the gaps of shaped castellation outlining fair agreement with the experiment. At the same time, a number of processes, like enhanced erosion of molten layer yet need to be implemented in the codes in order to increase the accuracy of predictions for ITER.

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“…Furthermore, as learned from the experiments on DIII-D, keeping the first optical element at elevated temperatures (~150°C) may also prolong the period over which the optical system throughput can be maintained in an acceptable range. 26 In the end, since already less than a handful of 30 min plasma discharges would, without any special measures taken, result in a first optical element contamination comparable to the one found after one year of operation in short pulse devices like JET, it will be unavoidable to frequently attempt in-situ cleaning of these components. With laser evaporation techniques and gas discharge cleaning not yet being sufficiently developed, only the method of heating the first plasma facing optical components to 300-400°C to remove a large fraction of the soft a-C:H layers overnight or over the weekend will be implemented in some of our diagnostics.…”

Section: Maintaining Optical Propertiesmentioning

confidence: 99%