Structure, tritium depth profile and desorption from 'plasma-facing' beryllium materials of ITER-Like-Wall at JET

Tritium depth profile and its temperature programmed desorption rate were determined for selected samples cut out of beryllium tiles removed from the Joint European Torus vacuum vessel during the 2012 shut down. A beryllium dissolution method under controlled conditions was used to determine the tritium depth profile in the samples, whereas temperature programmed desorption experiments were performed to assess tritium release pattern. Released tritium was measured using a proportional gas flow detector. Prior to desorption and dissolution experiments, the plasma-facing surfaces of the samples were studied by scanning electron microscopy and energy dispersive X-ray spectroscopy. Experimental results revealed that > 95% of the tritium was localized in the top 30 –45 μm of the ‘plasma-facing’ surface, however, possible tritium presence up to 100 μm cannot be excluded. During tem- perature programmed desorption at 4.8 K/min in the flow of purge gas He + 0.1% H 2 the tritium release started below 475 K, the most intense release occurred at 725 –915 K and the degree of detritiation of > 91% can be obtained upon reaching 1075 K. The total tritium activity in the samples was in range of 2 –32 kilo Becquerel per square centimetre of the plasma-facing surface area.EURATOM 63305

Comparison of the structure of the plasma-facing surface and tritium accumulation in beryllium tiles from JET ILW campaigns 2011-2012 and 2013-2014

In this study, beryllium tiles from Joint European Torus (JET) vacuum vessel wall were analysed and compared regarding their position in the vacuum vessel and differences in the exploitation conditions during two cam paigns of ITER-Like-Wall (LW) in 2011201211 WI) and 20132014 ILW 2) Tritium content in beryllium samples were assessed. Two methods were used to measure tritium content in the samples - dissolution under controlled conditions and tritium thermal desorption. Prior to desorption and dissolution experiments, scanning electron microscopy and energy dispersive x-ray spectroscopy were used to study structure and chemical composition of plasma.facing surface of the beryllium samples. Experimental results revealed that tritium content in the samples is in range of 2101 210 tritium atoms per square centimetre of the surface area with its highest content in the samples from the outer wall of the vacuum vessel (up to 1.910 atoms/cm in II.W1 campaign and 2410" atoms/cm in 11W2). The lowest content of tritium was found in the upper part of the vacuum vessel 2010 atoms/cm² and 2.010 atoms/cm in ILWI and IIW2, respectively). Results obtained from scanning electron microscopy has shown that surface morphology is different within single tile, however if to compare two campaigns main tendencies remains similarEURATOM 63305

Interpretative and predictive modelling of Joint European Torus collisionality scans

Transport modelling of Joint European Torus (JET) dimensionless collisionality scaling experiments in various operational scenarios is presented. Interpretative simulations at a fixed radial position are combined with predictive JETTO simulations of temperatures and densities, using the TGLF transport model. The model includes electromagnetic effects and collisions as well as □(→┬E ) X □(→┬B ) shear in Miller geometry. Focus is on particle transport and the role of the neutral beam injection (NBI) particle source for the density peaking. The experimental 3-point collisionality scans include L-mode, and H-mode (D and H and higher beta D plasma) plasmas in a total of 12 discharges. Experimental results presented in (Tala et al 2017 44th EPS Conf.) indicate that for the H-mode scans, the NBI particle source plays an important role for the density peaking, whereas for the L-mode scan, the influence of the particle source is small. In general, both the interpretative and predictive transport simulations support the experimental conclusions on the role of the NBI particle source for the 12 JET discharges

Comparison of the structure of the plasma-facing surface and tritium accumulation in beryllium tiles from JET ILW campaigns 2011-2012 and 2013-2014

In this study, beryllium tiles from Joint European Torus (JET) vacuum vessel wall were analysed and compared regarding their position in the vacuum vessel and differences in the exploitation conditions during two campaigns of ITER-Like-Wall (ILW) in 2011–2012 (ILW1) and 2013–2014 (ILW2) Tritium content in beryllium samples were assessed. Two methods were used to measure tritium content in the samples – dissolution under controlled conditions and tritium thermal desorption. Prior to desorption and dissolution experiments, scanning electron microscopy and energy dispersive x-ray spectroscopy were used to study structure and chemical composition of plasma-facing-surfaces of the beryllium samples. Experimental results revealed that tritium content in the samples is in range of 2·10$^{11}$–2·10$^{13}$ tritium atoms per square centimetre of the surface area with its highest content in the samples from the outer wall of the vacuum vessel (up to 1.9·10$^{13}$ atoms/cm$^{2}$ in ILW1 campaign and 2.4·10$^{13}$ atoms/cm$^{2}$ in ILW2). The lowest content of tritium was found in the upper part of the vacuum vessel (2.0·10$^{12}$ atoms/cm$^{2}$ and 2.0·10$^{11}$ atoms/cm$^{2}$ in ILW1 and ILW2, respectively). Results obtained from scanning electron microscopy has shown that surface morphology is different within single tile, however if to compare two campaigns main tendencies remains similar

Role of fast ion pressure in the isotope effect in JET L-mode plasmas

This paper presents results of JET ITER-like wall L-mode experiments in hydrogen and deuterium (D) plasmas, dedicated to the study of the isotope dependence of ion heat transport by determination of the ion critical gradient and stiffness by varying the ion cyclotron resonance heating power deposition. When no strong role of fast ions in the plasma core is expected, the main difference between the two isotope plasmas is determined by the plasma edge and the core behavior is consistent with a gyro-Bohm scaling. When the heating power (and the fast ion pressure) is increased, in addition to the difference in the edge region, also the plasma core shows substantial changes. The stabilization of ion heat transport by fast ions, clearly visible in D plasmas, appears to be weaker in H plasmas, resulting in a higher ion heat flux in H with apparent anti-gyro-Bohm mass scaling. The difference is found to be caused by the different fast ion pressure between H and D plasmas, related to the heating power settings and to the different fast ion slowing down time, and is completely accounted for in non-linear gyrokinetic simulations. The application of the TGLF quasi-linear model to this set of data is also discussed

COREDIV numerical simulation of high neutron rate JET-ILW DD pulses in view of extension to JET-ILW DT experiments

Two high performance JET-ILW pulses, pertaining to the 2016 experimental campaign, have been numerically simulated with the self-consistent code COREDIV with the aim of predicting the ELM-averaged power load to the target when extrapolated to DT plasmas. The input power of about 33 MW as well as the total radiated power and the average density are similar in the two pulses, but for one of them the density is provided by combined low gas puff and pellet injection, characterized by low SOL density, for the other one by gas fuelling only, at higher SOT. density. Considering the magnetic configuration of theses pulses and the presence of a significant amount of Ni (not included in the version of the code used for these simulations), a number of assumptions are made in order to reproduce numerically the main core and SOL experimental data. The extrapolation to DT plasmas at the original input power of 33 MW, and taking into account only the thermal component of the alpha-power, does not show any significant difference regarding the power to the target with respect to the DD case. In contrast, the simulations at auxiliary power 40 MW, both at the original I-p = 3 MA and at I-p = 4 MA, show that the power to the target for both pulses is possibly too high to be sustained for about 5 s by strike-point sweeping alone without any control by Ne seeding. Even though the target power load may decrease to about 13-15 MW with substantial Ne seeding for both pulses, as from numerical predictions, there are indications suggesting that the control of the power load may be more critical for the pulse with pellet injection, due to the reduced SOL radiation