IR reflectivity measurements depending on carbon film thickness

Abstract: In thermonuclear controlled fusion machines using magnetic confinement, carbonisations are realised to prevent metal impurities to enter into the fusion plasma made with hydrogen elements; it consists in helium glows in which methane gas is injected . The methane molecule is broken and the carbon deposits on all inside vessel surfaces : inner walls as well as optic elements like windows and mirrors . We studied the dependence of the reflectivity of infrared thermography stainless steel mirrors with carbon films thickness in the 3-5 m bandwidth . The presented results show a decrease of less than 10% of the temperature announced by the camera . . Magnetic fusion : Thermonuclear controlled fusion Although plasma particles are trapped in non-material magnetic barriers, atomic collisions appear and diffusion, convection,conduction and radiation phenomena occur; this implies heating of contact surfaces . That is why an infrared thermographic system is developped to measure and survey these heatings which can lead to a failure of one of the plasma facing component inside the tokamak [2] . . IR thermographic diagnostic : The Tore Supra IR thermographic diagnostic consists, until now, in three endoscopes [fig 2] situated at 120° one from the other on the top of the tokamak; they are remotely controled so that every element inside the vessel can be observed . Images of objects inside the vessel are captured by a movable stainless steel mirror and returned to an optical system through a sapphire window . This window is necessary to keep http://dx.doi.org/10.21611/qirt.2000.049 ultra high vacuum inside the torus vessel . Inframetrics 3-5 m cameras recuperate images from the optical system which transported them . Plasma inside the tokamak encounters surfaces made of, or covered with graphite . This graphite is sputtered and redeposited during or after plasma is stopped . Furthermore carbonisations are made to cover metal inner walls of the vessel with carbon so that metal impurities rate inside the plasma is lower . Carbonisation, and its erosion, was extensively studied . Experimental apparatus : Our carbonisation system was installed in a special vessel Two 2 polished mirrors, set back to back, are supposed to be hidden from carbon deposition and centimeter by centimeter are exposed to the glow (see . Experimental results : Six carbon films 1 cm x 3 cm wide were obtained : 50 monolayers, 150, 250, 350, 450 and 550 monolayers thick [ The two mirrors present the same visual aspect . A correlation between film thickness and colour was established by J. Winter in his study of carbonisation in Textor tokamak Reflectivity results for both mirrors are presented below [ Reproducibility of our reflectivity measurements is in the order of +/-2 % . Our deposits need special care because of their poor adhesion : we lost part of the thickest deposits by rubbing with a plastic bag . Mirrors surfaces were made very smooth and then it is not so surprising that our films have little adhesion as thick they are . Furthermore, electrostatic interaction may increase this effect . Indeed carbonisation films have quite high electrical resistivity . For example, on Textor samples 1 -10 6 cm was typically measured Measurement of the 550 monolayers zone where the deposit is and where it peeled, was undertaken . Reflectivity of the undamaged part of the film is the lowest . Reflectivity of the peeled film is half between that latter and that of 0 -150 monolayers . If we consider the worst result which means 550 undamaged monolayers deposited, we find that the real temperature is underestimated by less than 10 per cent; this may be dangerous in our application especially at high temperatures : for example, the copper melting point is at 1083°C, with our results the temperature indicated by infrared cameras is around 975°C, but inside the copper 15 bar water circulate to refregirate the plasma facing components which means that with our measurement we wouldn't understand why there is a water leak in the machine; this leak implies no experiment for three months to mend damaged parts and to restart the machine . . Conclusion : We realised carbonisation with variable thicknessses on two polished stainless steel mirrors used in the infrared 3-5 m band . Their reflectivity before and after carbonisation was measured, on the different thicknesses of the films formed . Below 150 monolayers the reflectivity does not change . Beyond 150 monolayers, a decrease is observed . If deposit peels, reflectivity is only modified but does not come back to the value with few deposit . Over 450 monolayers the deposit shows waves which show poor film adhesion . The two mirrors have the same results even if they were set back to back for carbonisation, one looking at the anode, the other looking at the opposite . We deduced from these reflectivity measurements that in the case of the thickest deposit the real temperature of the tokamak element is under evaluated by less than 10 % . Acknowledgements : The authors want to thank A Grosman for fruitfull discussion and P. Maillet and his team for their advices and help for that work to be done

Velocity-space sensitivity of the time-of-flight neutron spectrometer at JET

The velocity-space sensitivities of fast-ion diagnostics are often described by so-called weight functions. Recently, we formulated weight functions showing the velocity-space sensitivity of the often dominant beam-target part of neutron energy spectra. These weight functions for neutron emission spectrometry (NES) are independent of the particular NES diagnostic. Here we apply these NES weight functions to the time-of-flight spectrometer TOFOR at JET. By taking the instrumental response function of TOFOR into account, we calculate time-of-flight NES weight functions that enable us to directly determine the velocity-space sensitivity of a given part of a measured time-of-flight spectrum from TOFOR

WEST full tungsten operation with an ITER grade divertor

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction

Operating a full tungsten actively cooled tokamak: overview of WEST first phase of operation

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017–2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m−2 were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described

Model for screening of resonant magnetic perturbations by plasma in a realistic tokamak geometry and its impact on divertor strike points

This work addresses the question of the relation between strike-point splitting and magnetic stochasticity at the edge of a poloidally diverted tokamak in the presence of externally imposed magnetic perturbations. More specifically, ad-hoc helical current sheets are introduced in order to mimic a hypothetical screening of the external resonant magnetic perturbations by the plasma. These current sheets, which suppress magnetic islands, are found to reduce the amount of splitting expected at the target, which suggests that screening effects should be observable experimentally. Multiple screening current sheets reinforce each other, i.e. less current relative to the case of only one current sheet is required to screen the perturbation.Comment: Accepted in the Proceedings of the 19th International Conference on Plasma Surface Interactions, to be published in Journal of Nuclear Materials. Version 2: minor formatting and text improvements, more results mentioned in the conclusion and abstrac

Relationship of edge localized mode burst times with divertor flux loop signal phase in JET

A phase relationship is identified between sequential edge localized modes (ELMs) occurrence times in a set of H-mode tokamak plasmas to the voltage measured in full flux azimuthal loops in the divertor region. We focus on plasmas in the Joint European Torus where a steady H-mode is sustained over several seconds, during which ELMs are observed in the Be II emission at the divertor. The ELMs analysed arise from intrinsic ELMing, in that there is no deliberate intent to control the ELMing process by external means. We use ELM timings derived from the Be II signal to perform direct time domain analysis of the full flux loop VLD2 and VLD3 signals, which provide a high cadence global measurement proportional to the voltage induced by changes in poloidal magnetic flux. Specifically, we examine how the time interval between pairs of successive ELMs is linked to the time-evolving phase of the full flux loop signals. Each ELM produces a clear early pulse in the full flux loop signals, whose peak time is used to condition our analysis. The arrival time of the following ELM, relative to this pulse, is found to fall into one of two categories: (i) prompt ELMs, which are directly paced by the initial response seen in the flux loop signals; and (ii) all other ELMs, which occur after the initial response of the full flux loop signals has decayed in amplitude. The times at which ELMs in category (ii) occur, relative to the first ELM of the pair, are clustered at times when the instantaneous phase of the full flux loop signal is close to its value at the time of the first ELM