Substrate-dependent deposition efficiency on mirrors exposed in the TCV divertor

Given the planned extensive use of metallic mirrors in the optical diagnostics systems of ITER, the study of reflectivity changes induced by erosion and/or redeposition of impurities on the mirror surfaces is of primary importance for the reliability of the diagnostics signals. This contribution will demonstrate that the mirror material choice can exert a significant influence on the relative importance of erosion/redeposition affecting the mirror reflectivity. A dedicated manipulator has been designed to allow exposure of mirror samples in the divertor region of the Tokamak à Configuration Variable (TCV) tokamak. Mirrors from different materials have been exposed both during short experimental campaigns and boronisation procedures. Before and after exposures the mirrors were characterized with different surface analysis techniques (XPS, SIMS, SEM, EDX, ellipsometry). Under identical exposure conditions, the mirror material can strongly influence the deposit thickness found on the sample: the carbon layer thickness on a Si sample is much higher than on a Mo sample. These results have potentially important consequences for the first mirror material choice in ITER.     

Plasma and antenna coupling characterization in ICRF-wall conditioning experiments

Ion Cyclotron Wall Conditioning (ICWC) discharges, in pulsed-mode operation, were carried out in the divertor tokamaks ASDEX Upgrade (AUG) and JET to simulate the scenario of ITER wall conditioning at half-field (AUG) and full-field (JET). ICWC-plasma and antenna coupling characterization results obtained during the Ion Cyclotron Resonance Frequency (ICRF)-Wall Conditioning experiments performed in helium-hydrogen mixture in AUG and helium-deuterium mixtures in JET are presented here. Safe operational regimes for optimum ICWC in ITER could be explored for different magnetic fields. Satisfactory antenna coupling in the Mode Conversion scenario along with reproducible generation of ICRF plasmas and reliable wall conditioning were achieved by coupling RF power from one or two ICRF antennas at two (AUG, JET) different resonant frequencies. These results are in qualitative agreement with the predictions of 1-D TOMCAT code. Present study of ICWC indicates towards the beneficial effect of application of an additional (along with toroidal magnetic field) stationary vertical (B_V ? B_T) magnetic field on antenna coupling and plasma parameters. The results obtained from JET and AUG tokamaks, presented in this paper, emphasizes the proposed phenomenological schemes for further development of ICWC in superconducting tokamaks.     

Blister formation on tungsten damaged by high energy particle irradiation

In order to investigate the effect of radiation damage on hydrogen behavior in tungsten, tungsten samples with radiation damage of up to 3.5 dpa were irradiated by a mixed hydrogen-carbon ion beam. The radiation damage was produced with 700 keV negative hydrogen ion beam irradiation. The number density of blisters produced by the mixed ion beam irradiation decreased with increasing radiation damage. This was especially observed for blisters with diameters of 20 μm or less. This result showed that radiation damage produced by high-energy particle irradiation suppresses blister formation on tungsten surfaces.     

Finalizing the ITER divertor design: The key role of SOLPS modeling

The paper presents a review of the development of edge plasma modeling at ITER and of its interaction with the evolving divertor design. The SOLPS (B2-Eirene) code has been developed for, and applied to, the evaluation and the design of the ITER divertor for the last 15 years. With respect to the physics and engineering design, divertor modeling had started as an evaluation tool and has developed into essential design tool synthesizing information from theoretical analysis, experimental studies, and engineering intuition. Examples given in the paper illustrate this process.     

On the carbon and tungsten sputtering rate in a magnetron discharge

Magnetron discharge as sputtering source can serve as an alternative tool for the study of the plasma-wall interaction, with applications for ITER divertor. The present work reports on the influence of the target power density and the nature of the projectile on the erosion of C and W targets. The experimental results concern the sputtering rate of carbon and tungsten targets of a d.c. magnetron discharge in argon and helium atmosphere, at different gas pressures in the range of 10-100 mTorr and discharge power densities up to 40 W cm⁻² while the discharge current intensity was used as control parameter. In this investigation, carbon and tungsten sputtering rates were measured using two conventional methods based on gravimetric mass loss and profilometry. Target erosion profiles were compared with the profiles of the ion energy flux bombarding the target, calculated from a 2D fluid model.     

Tokamak field error measurements with an electron beam in KSTAR

It is possible to detect the presence of small field errors in a tokamak with an electron beam. This was demonstrated earlier on T-15 and TEXTOR. This paper discusses the concept, past experience on these tokamaks, calculations for the Korea Superconducting Tokamak Advanced Research (KSTAR) device, an electron beam source, measurement devices for these measurements, and some results. It is shown that small toroidally averaged field errors can be detected by this method. A low voltage electron beam (e-beam) gun and fluorescent screen were mounted in a vertical port and inserted into the vacuum vessel at the end of the KSTAR 2nd campaign plasma experiments. A camera with a narrow field of view was mounted in midplane port in a tube tangent to the field lines at R ∼ 1.3 m and photographed the beam striking the screen. The poloidal field (PF) currents were held constant during the camera exposure period. Many shots with various PF coils energized were made and the deflections of the e-beam were measured. The measurements were made with a camera integration time of 300 ms because of the low light intensity. The results show that there are large field errors that diminish as the PF currents are raised. There appears to be no significant up-down asymmetry for static fields. Measurements with a 7 PF coil scenario with a calculated field null located at e-beam radial position show much larger fields than calculated. KSTAR was constructed with Incoloy 908 conduit using cable-in-conduit conductors (CICC) in 10 of the 14 PF coils and all 16 of the toroidal field (TF) coils. Incoloy 908 has a relative magnetic permeability, μ, of about 10. The field errors appear to be largely due to Incoloy 908.     

Heat load behaviors of plasma sprayed tungsten coatings on copper alloys with different compliant layers

Plasma sprayed tungsten (PS-W) coatings with the compliant layers of titanium (Ti), nickel-chromium-aluminum (NiCrAl) alloys and W/Cu mixtures were fabricated on copper alloys, and their properties of the porosity, oxygen content, thermal conductivity and bonding strength were measured. High heat flux tests of actively cooled W coatings were performed by means of an electron beam facility. The results indicated that APS-W coating showed a poorer heat transfer capability and thermo-mechanical properties than VPS-W coating, and the compliant layers improved W coating performance under the heat flux load. Among three compliant layers, W/Cu was the preferable because of its better effects on heat removal and stress alleviating. The optimization of W/Cu compliant layer found that 0.1 mm and 25 vol.%W was optimum compliant layer structure for 1 mm W coating, which induced a 23% reduction of the maximum stress compared to the sharp interface, and the plastic strain was reduced to 0.01% from 1.55%.     

Thermo-mechanical calculations on operation temperature limits of tungsten as plasma facing material

Tungsten is a candidate material as a plasma facing material in the next step fusion devices. The material surface will be exposed to transient heat loads as well as steady-state heat loads. The present work describes the thermo-mechanical analysis of tungsten by finite element calculation. It is shown that tungsten has a strict operational temperature limit under transient heat loads. For the ITER-grade W, the operation limit of the base surface-temperature was calculated to be in a range of 400-780 °C under an applied transient heat load of 0.2 GW/m² for 0.5 ms in order to avoid plastic deformation of W.     

Deuterium inventory in plasma facing materials by means of NRA: A microbeam probe approach

Particle retention in tokamak walls is a key issue for long time discharges in future thermonuclear fusion reactors. Plasma wall interactions drive the fuel retention through two major mechanisms: co-deposition with carbon produced by wall erosion and particle retention in wall materials. In this study, we report results obtained from the tokamak Tore Supra, from which two types of samples were analyzed by means of micro-NRA: (i) small pieces of deposited carbon layers were collected after cumulative discharges and deuterium contents were measured; (ii) carbon fiber composite (CFC) samples, immersed in the plasma during an experimental campaign were also analyzed. 3D deuterium elemental mapping demonstrated that deuterium can be trapped at depths much higher than usual implantation depths and deep local retention sites have been evidenced and localized.This study demonstrates that μNRA can be used for assessment of deuterium post-mortem inventory in tokamaks, both by measuring uniformly distributed deuterium in small fragments of deposited carbon layers and by locally describing deuterium 2D and 3D distributions in complex structures.     

Design and experimental results of feedback control of Ohmic-heating transformer magnetic flux by LHCD power in HT-7 Tokamak

In order to make a research on long pulse or even steady state operation with non-inductive drive in plasma discharge, a new feedback control scheme instead of the previous one has been designed and operated in HT-7 [HT-7 team presented by J. Li, et al., Plasma Phys. Control. Fusion 42 (2) (2000) 135-146] Tokamak experiment, 2004. Consumption of iron-core transformer magnetic flux (MFT) is feedback controlled for the first time by power of lower hybrid current drive (LHCD) P_LH, when the Ohmic-heating circuit current can maintain the plasma current I_P constant with another feedback control loop, which make MFT evolve at alternating-change state to avoid flux saturation. Plasma current I_P can be maintained steadily up to 120 s in this operation mode at reduced plasma parameters (I_P ≈ 50-100 KA, average density , P_LH = 100-200 KW). Design and experimental results are presented in the paper, which including control model analysis, configurations of control system and MFT feedback control experiments in HT-7. The high voltage power supply (HVPS) of LHCD is the main controller that regulates the LHCD power into the plasma to control the MFT.     

Thermal properties of redeposition layers in the JT-60U divertor region

Thermal properties of the redeposition layer on the inner plate of the W-shaped divertor of JT-60U have been measured with laser flash method so as to estimate transient heat loads onto the divertor. Morphology analysis of the redeposition layer was conducted with a scanning electron microscope. Measurement of a redeposition layer sample of more than 200 μm thick, which had been produced near the most frequent striking point, showed following results: (1) the bulk density of the redeposition layer is about half of that of carbon fiber composite material; (2) the specific heat of the layer is roughly equal to that of the isotropic graphite; (3) the thermal conductivity of the redeposition layer is two orders of magnitude smaller than that of the carbon fiber composite. This low thermal conductivity of the redeposition layer is considered to be caused by a low graphitization degree of the redeposition layer. The difference between the divertor heat loads and the loss of the plasma stored energy becomes smaller taking account of thermal properties of the redeposition layer on the inner divertor, whereas estimated heat loads due to the ELMs is still larger than the loss. This is probably caused by the poloidal distribution of the thermal properties.