| 1. |
- Bernert, M., et al.
(författare)
-
Power exhaust by SOL and pedestal radiation at ASDEX Upgrade and JET
- 2017
-
Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 12, s. 111-118
-
Tidskriftsartikel (refereegranskat)abstract
- Future fusion reactors require a safe, steady state divertor operation. A possible solution for the power exhaust challenge is the detached divertor operation in scenarios with high radiated power fractions. The radiation can be increased by seeding impurities, such as N for dominant scrape-off-layer radiation, Ne or Ar for SOL and pedestal radiation and Kr for dominant core radiation. Recent experiments on two of the all-metal tokamaks, ASDEX Upgrade (AUG) and JET, demonstrate operation with high radiated power fractions and a fully-detached divertor by N, Ne or Kr seeding with a conventional divertor in a vertical target geometry. For both devices similar observations can be made. In the scenarios with the highest radiated power fraction, the dominant radiation originates from the confined region, in the case of N and Ne seeding concentrated in a region close to the X-point. Applying these seed impurities for highly radiative scenarios impacts local plasma parameters and alters the impurity transport in the pedestal region. Thus, plasma confinement and stability can be affected. A proper understanding of the effects by these impurities is required in order to predict the applicability of such scenarios for future devices.
|
|
| 2. |
- Fortuna-Zalesna, E., et al.
(författare)
-
Studies of dust from JET with the ITER-Like Wall : Composition and internal structure
- 2017
-
Ingår i: Nuclear Materials and Energy. - : ELSEVIER. - 2352-1791. ; 12, s. 582-587
-
Tidskriftsartikel (refereegranskat)abstract
- Results are presented for the dust survey performed at JET after the second experimental campaign with the ITER-Like Wall: 2013-2014. Samples were collected on adhesive stickers from several different positions in the divertor both on the tiles and on the divertor carrier. Brittle dust-forming deposits on test mirrors from the inner divertor wall were also studied. Comprehensive characterization accomplished by a wide range of high-resolution microscopy techniques, including focused ion beam, has led to the identification of several classes of particles: (i) beryllium flakes originating either from the Be coatings from the inner wall cladding or Be-rich mixed co-deposits resulting from material migration; (ii) beryllium droplets and splashes; (iii) tungsten and nickel-rich (from Inconel) droplets; (iv) mixed material layers with a various content of small (8-200 nm) W-Mo and Ni-based debris. A significant content of nitrogen from plasma edge cooling has been identified in all types of co-deposits. A comparison between particles collected after the first and second experimental campaign is also presented and discussed. (C) 2016 Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND license.
|
|
| 3. |
- Heinola, K., et al.
(författare)
-
Long-term fuel retention and release in JET ITER-Like Wall at ITER-relevant baking temperatures
- 2017
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 57:8
-
Tidskriftsartikel (refereegranskat)abstract
- The fuel outgassing efficiency from plasma-facing components exposed in JET-ILW has been studied at ITER-relevant baking temperatures. Samples retrieved from the W divertor and Be main chamber were annealed at 350 and 240 degrees C, respectively. Annealing was performed with thermal desoprtion spectrometry (TDS) for 0, 5 and 15 h to study the deuterium removal effectiveness at the nominal baking temperatures. The remained fraction was determined by emptying the samples fully of deuterium by heating W and Be samples up to 1000 and 775 degrees C, respectively. Results showed the deposits in the divertor having an increasing effect to the remaining retention at temperatures above baking. Highest remaining fractions 54 and 87% were observed with deposit thicknesses of 10 and 40 mu m, respectively. Substantially high fractions were obtained in the main chamber samples from the deposit-free erosion zone of the limiter midplane, in which the dominant fuel retention mechanism is via implantation: 15 h annealing resulted in retained deuterium higher than 90%. TDS results from the divertor were simulated with TMAP7 calculations. The spectra were modelled with three deuterium activation energies resulting in good agreement with the experiments.
|
|
| 4. |
- Joffrin, E., et al.
(författare)
-
Impact of divertor geometry on H-mode confinement in the JET metallic wall
- 2017
-
Ingår i: Nuclear Fusion. - : Institute of Physics Publishing (IOPP). - 0029-5515 .- 1741-4326. ; 57:8
-
Tidskriftsartikel (refereegranskat)abstract
- Recent experiments with the ITER-like wall have demonstrated that changes in divertor strike point position are correlated with strong modification of the global energy confinement. The impact on energy confinement is observable both on the pedestal confinement and core normalised gradients. The corner configuration shows an increased core density gradient length and ion pressure indicating a better ion confinement. The study of neutral re-circulation indicates the neutral pressure in the main chamber varies inversely with the energy confinement and a correlation between the pedestal total pressure and the neutral pressure in the main chamber can be established. It does not appear that charge exchange losses nor momentum losses could explain this effect, but it may be that changes in edge electric potential are playing a role at the plasma edge. This study emphasizes the importance of the scrape-off layer (SOL) conditions on the pedestal and core confinement.
|
|
| 5. |
- Koechl, F., et al.
(författare)
-
Modelling of transitions between L- and H-mode in JET high plasma current plasmas and application to ITER scenarios including tungsten behaviour
- 2017
-
Ingår i: Nuclear Fusion. - : IOP PUBLISHING LTD. - 0029-5515 .- 1741-4326. ; 57:8
-
Tidskriftsartikel (refereegranskat)abstract
- The dynamics for the transition from L-mode to a stationary high QDT H-mode regime in ITER is expected to be qualitatively different to present experiments. Differences may be caused by a low fuelling efficiency of recycling neutrals, that influence the post transition plasma density evolution on the one hand. On the other hand, the effect of the plasma density evolution itself both on the alpha heating power and the edge power flow required to sustain the H-mode confinement itself needs to be considered. This paper presents results of modelling studies of the transition to stationary high QDT H-mode regime in ITER with the JINTRAC suite of codes, which include optimisation of the plasma density evolution to ensure a robust achievement of high QDT regimes in ITER on the one hand and the avoidance of tungsten accumulation in this transient phase on the other hand. As a first step, the JINTRAC integrated models have been validated in fully predictive simulations (excluding core momentum transport which is prescribed) against core, pedestal and divertor plasma measurements in JET C-wall experiments for the transition from L-mode to stationary H-mode in partially ITER relevant conditions (highest achievable current and power, H-98,H-y similar to 1.0, low collisionality, comparable evolution in P-net/PL-H, but different rho(*), T-i/T-e, Mach number and plasma composition compared to ITER expectations). The selection of transport models (core: NCLASS + Bohm/gyroBohm in L-mode/GLF23 in H-mode) was determined by a trade-off between model complexity and efficiency. Good agreement between code predictions and measured plasma parameters is obtained if anomalous heat and particle transport in the edge transport barrier are assumed to be reduced at different rates with increasing edge power flow normalised to the H-mode threshold; in particular the increase in edge plasma density is dominated by this edge transport reduction as the calculated neutral influx across the separatrix remains unchanged (or even slightly decreases) following the H-mode transition. JINTRAC modelling of H-mode transitions for the ITER 15 MA/5.3 T high Q(DT) scenarios with the same modelling assumptions as those being derived from JET experiments has been carried out. The modelling finds that it is possible to access high Q(DT) conditions robustly for additional heating power levels of P-AUX >= 53 MW by optimising core and edge plasma fuelling in the transition from L-mode to high Q(DT) H-mode. An initial period of low plasma density, in which the plasma accesses the H-mode regime and the alpha heating power increases, needs to be considered after the start of the additional heating, which is then followed by a slow density ramp. Both the duration of the low density phase and the density ramp-rate depend on boundary and operational conditions and can be optimised to minimise the resistive flux consumption in this transition phase. The modelling also shows that fuelling schemes optimised for a robust access to high Q(DT) H-mode in ITER are also optimum for the prevention of the contamination of the core plasma by tungsten during this phase.
|
|
| 6. |
- Lahtinen, A., et al.
(författare)
-
Deuterium retention in the divertor tiles of JET ITER-Like wall
- 2017
-
Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 12, s. 655-661
-
Tidskriftsartikel (refereegranskat)abstract
- Divertor tiles removed after the second JET ITER-Like Wall campaign 2013-2014 (ILW-2) were studied using Secondary Ion Mass Spectrometry (SIMS). Measurements show that the thickest beryllium (Be) dominated deposition layers are located at the upper part of the inner divertor and are up to similar to 40 mu m thick at the lower part of Tile 0 exposed in 2011-2014. The highest deuterium (D) amounts (>8 . 10 18 at./cm(2)), in contrast, were found on the upper part of Tile 1 (2013-2014), where the Be deposits are similar to 10 mu m thick. D was mainly retained in the near-surface layer of the Be deposits but also deeper in tungsten (W) and molybdenum (Mo) layers of the marker coated tiles, especially at W-Mo layer interfaces. D retention for the ILW-2 divertor tiles is higher than for the first campaign 2011-2012 (ILW-1) and probable reasons for the difference are that SIMS measurements for the ILW-2 samples were done deeper than for the ILW-1 samples, some of the tiles were exposed during both ILW-1 and ILW-2 and therefore had a longer exposure time, and the differences between ILW-1 and ILW-2 campaigns e.g. in strike point distributions and injected powers.
|
|
| 7. |
- Moser, L., et al.
(författare)
-
Investigation and plasma cleaning of first mirrors coated with relevant ITER contaminants : beryllium and tungsten
- 2017
-
Ingår i: Nuclear Fusion. - : IOP PUBLISHING LTD. - 0029-5515 .- 1741-4326. ; 57:8
-
Tidskriftsartikel (refereegranskat)abstract
- In order to extend the investigation of the plasma cleaning of ITER first mirrors, a set of molybdenum mirrors was coated in a laboratory with ITER-relevant contaminants, namely beryllium and tungsten. Different coating techniques as well as several contaminant compositions were used to ensure a large variety of films to clean, completing a previous study conducted on mirrors exposed in the JET ITER-like wall (tokamak deposits) [ 1]. Due to the toxicity of beryllium, the samples were treated in a vacuum chamber specially built for this purpose. The cleaning was performed using capacitively coupled RF plasma and evaluated by performing reflectivity measurements, scanning electron microscopy, x-ray photoelectron spectroscopy and ion beam analysis. The removal of all types of contaminants was achieved by using different plasma compositions (argon, helium and mixtures of the two) with various ion energies (from 200-600 eV) and in some cases the mirror's reflectivity was restored towards initial values. Pure helium discharges were capable of removing mixed beryllium/tungsten layers and oxidized molybdenum. In addition, no significant increase in the diffuse reflectivity of the mirrors was observed for the helium cleaning, though this was the case for some samples cleaned with argon. Helium is therefore appropriate for cleaning all mirrors in ITER leading to a possible cleaning regime where the entire vessel is filled with He and all mirrors are cleaned simultaneously without damaging their surfaces.
|
|
| 8. |
- Telesca, G., et al.
(författare)
-
High power neon seeded JET discharges : Experiments and simulations
- 2017
-
Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 12, s. 882-886
-
Tidskriftsartikel (refereegranskat)abstract
- A series of neon seeded JET ELMy H-mode pulses is considered from the modeling as well as from the experimental point of view. For two different Ne seeding rates and two different D puffing gas levels the heating power, P-heat, is in the range 22-29.5 MW. The main focus is on the numerical reconstruction of the total radiated power (which mostly depends on the W concentration) and its distribution between core and divertor and of Z(eff)(which mostly depends on the Ne concentration). To model self-consistently the core and the SOL two input parameters had to be adjusted case by case: the SOL diffusivity, D SOL, and the core impurity inward pinch, v(pinch). D-SOL had to be increased with increasing Gamma(Ne) and the level of v(pinch) had to be changed, for any given Gamma(Ne), according to the level of P-heat : it decreases with increasing P-heat. Since the ELM frequency, f(ELM), is experimentally correlated with P-heat, (it increases with P-heat) the impurity inward pinch can be seen as to depend on f(ELM). Therefore, to maintain a low v(pinch) level (i.e. high f(ELM)) Gamma(Ne)/P-heat should not exceed a certain threshold, which slightly increases with the Gamma(D) puffing rate. This might lead to a limitation in the viability of reducing the target heat load by Ne seeding at moderate Gamma(D), while keeping Z(eff) at acceptably low level.
|
|
