Showing papers in "Nuclear materials and energy in 2017"
TL;DR: In this article, the effect of edge power loading on the shape of the ITER divertor was investigated, and it was concluded that the geometrical approximation for leading edge power load on radially misaligned poloidal leading edges is indeed valid.
Abstract: The key remaining physics design issue for the ITER tungsten (W) divertor is the question of monoblock (MB) front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3 mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM) or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA) and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets.
120 citations
TL;DR: A newly established scaling of the ELM energy fluence using dedicated data sets from JET operation with CFC & ILW plasma facing components (PFCs), ASDEX Upgrade (AUG) operation with both CFC an...
Abstract: A newly established scaling of the ELM energy fluence using dedicated data sets from JET operation with CFC & ILW plasma facing components (PFCs), ASDEX Upgrade (AUG) operation with both CFC an ...
100 citations
TL;DR: In this article, a possible solution for the power exhaust challenge is the detached divertor operation in scenarios with high radiated power in a future fusion nuclear power plant, where future fusion reactors require a safe, steady state divertor.
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 ...
89 citations
TL;DR: In this article, a detailed match of the high field side scrape-off layer plasma is not only important for local plasma parameters, but can lead to strong changes in global parameters.
Abstract: The understanding of divertor physics and the evolution of divertor detachment is crucial for developing the capability to model power exhaust in current experiments and reliably predict it for future fusion devices. In simulations of ASDEX Upgrade, an experimentally observed region of high density in the high field side scrape-off layer has been recovered. Validated modeling with SOLPS5.0 shows that a detailed match of the high field side scrape-off layer plasma is not only important for local plasma parameters, but can lead to strong changes in global parameters. Drifts play a crucial role in lower-single null discharges with forward toroidal field (∇B→-drift pointing down). Their inclusion changes the spatial extent as well as the radial and poloidal gradients of the high field side high density. Adapted transport coefficients that take into account core fueling by plasma diffusion due to the presence of the high field side density and drift-driven radial fluxes now reconcile the modeled deuterium compression ratio, divertor neutral density, neutral radiation levels and deuterium fueling rates with experimental measurements. The onset of strong volume recombination in the simulations now allows to remove the previously necessary increase of perpendicular transport in the inner divertor from the simulations.
83 citations
TL;DR: In this paper, the status of the literature for several thermophysical properties of pure solid and liquid tungsten which constitute input for the modelling of intense plasma-surface interaction pheno...
Abstract: The status of the literature is reviewed for several thermophysical properties of pure solid and liquid tungsten which constitute input for the modelling of intense plasma-surface interaction pheno ...
75 citations
TL;DR: A pulsed laser-induced desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER as mentioned in this paper.
Abstract: As a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1 kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513 K for the FW and 623 K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.
59 citations
TL;DR: In this article, polycrystalline tungsten was self-damaged by 20 MeV Tungsten ions up to a calculated damage dose in the damage peak of 0.23 dpa.
Abstract: Recrystallized, polycrystalline tungsten was self-damaged by 20 MeV tungsten ions up to a calculated damage dose in the damage peak of 0.23 dpa. The time to acquire this dose and hence the average damaging dose rate was varied from 6 × 10 -3 to 4 × 10 -6 dpa/s, the latter coming close to the damage dose rate expected from fusion neutrons in future devices such as ITER and DEMO. One series was conducted at 295 K and one at 800 K to check for possible effects of defect evolution at elevated temperature. The created damage was decorated afterwards with a deuterium plasma at low ion energy of 19 D/m 2 until saturation to derive a measure for the defect density that can retain hydrogen isotopes. 3 He nuclear reaction analysis (NRA) was applied to derive the deuterium depth profile and the maximum concentration in the damage peak. Neither for the 295 K nor for the 800 K series a variation in deuterium retention with damage dose rate was found.
50 citations
TL;DR: In this article, an activity to characterize W monoblock materials was launched at the IO to understand correlation between the macro-cracks appearance and W properties, and the outcome highlighted that the higher the recrystallization resistance, the lower the number of cracks detected during high heat flux tests.
Abstract: In the full-tungsten divertor qualification program at ITER Organization, macro-cracks, so called self-castellation were found in a fraction of tungsten monoblocks during cyclic high heat flux loading at 20MW/m 2 . The number of monoblocks with macro-cracks varied with the tungsten products used as armour material. In order to understand correlation between the macro-crack appearance and W properties, an activity to characterize W monoblock materials was launched at the IO. The outcome highlighted that the higher the recrystallization resistance, the lower the number of cracks detected during high heat flux tests. Thermo-mechanical finite element modelling demonstrated that the maximum surface temperature ranges from 1800 °C to 2200 °C and in this range recrystallization of tungsten occurred. Furthermore, it indicated that loss of strength due to recrystallization is responsible for the development of macro-cracks in the tungsten monoblock.
48 citations
TL;DR: In this paper, a new analysis of components removed following the second JET ITER-like wall campaign 2013-14 concentrating on the upper inner divertor, inner and outer divertor corners, was presented.
Abstract: The work presented draws on new analysis of components removed following the second JET ITER-like wall campaign 2013-14 concentrating on the upper inner divertor, inner and outer divertor corners, ...
47 citations
TL;DR: In this paper, a simple model of detachment is developed to evaluate the required upstream density, based on further taking into account dynamic pressure balance, and a remarkable general result is found that the upstream density divided by the Greenwald-limit density scales as n up /n GW ∝ (P 5/8 /B 3/8 ) T det 1/2 /(e cool ǫ+γ T det ), with no explicit size scaling.
Abstract: Fusion power plants are likely to require near complete detachment of the divertor plasma from the divertor target plates, in order to have both acceptable heat flux at the target to avoid prompt damage and also acceptable plasma temperature at the target surface, to minimize long-term erosion. However hydrogenic and impurity puffing experiments show that detached operation leads easily to x-point MARFEs, impure plasmas, degradation in confinement, and lower helium pressure at the exhaust. The concept of the Lithium Vapor Box Divertor is to use local evaporation and strong differential pumping through condensation to localize low-Z gas-phase material that absorbs the plasma heat flux and so achieve detachment while avoiding these difficulties. The vapor localization has been confirmed using preliminary Navier–Stokes calculations. We use ADAS calculations of e cool , the plasma energy lost per injected lithium atom, to estimate the lithium vapor pressure, and so temperature, required for detachment, taking into account power balance. We also develop a simple model of detachment to evaluate the required upstream density, based on further taking into account dynamic pressure balance. A remarkable general result is found, not just for lithium-vapor-induced detachment, that the upstream density divided by the Greenwald-limit density scales as n up /n GW ∝ ( P 5/8 / B 3/8 ) T det 1/2 /(e cool + γ T det ), with no explicit size scaling. T det is the temperature just before strong pressure loss, assumed to be ∼ ½ of the ionization potential of the dominant recycling species, and γ is the sheath heat transmission factor.
46 citations
TL;DR: In this article, it is shown how the shape or position of the density profile in the core and the position of edge pressure gradient influences global confinement of H-mode plasmas when the direction is inwards.
Abstract: Plasmas in machines with all metal plasma facing components have a lower Zeff, less radiation cooling in the scrape-off layer and divertor regions and are prone to impurity accumulation in the core. Higher gas puff and the seeding of low-Z impurities are applied to prevent impurity accumulation, to increase the frequency of edge localised modes and to cool the divertor. A lower power threshold for the transition from low-confinement mode to high confinement mode has been found in all metal wall machines when compared to carbon wall machines. The application of lithium before or during discharges can lead to ELM free H-modes. The seeding of high-Z impurities increases core radiation, reduces the power flux across the separatrix and, if applied in the right amount, does not lead to deterioration of the confinement. All these effects have in common that they can often be explained by the shape or position of the density profile. Not only the peakedness of the density profile in the core but also the position of the edge pressure gradient influences global confinement. It is shown how (i) ionisation in the pedestal region due to higher reflection of deuterium from high-Z walls, (ii) reduced recycling in consequence of lithium wall conditioning, (iii) the fostering of edge modes with lithium dropping, (iv) increased gas puff and (v) the cooling of the scrape-off layer by medium-Z impurities such as nitrogen affect the edge density profile. The consequence is a shift in the pressure profile relative to the separatrix, leading to improved pedestal stability of H-mode plasmas when the direction is inwards.
TL;DR: In this paper, the annealing influence on the yttrium oxide microstructure is investigated, and the results show that the results represent a mixture of different crystal structures including the metastable monoclinic phase and the stable cubic phase.
Abstract: Yttrium oxide thin films were prepared by reactive magnetron sputtering in different deposition condition with various oxygen flow rates. The annealing influence on the yttrium oxide film microstructure is investigated. The oxygen flow shows a hysteresis behavior on the deposition rate. With a low oxygen flow rate, the so called metallic mode process with a high deposition rate (up to 1.4 µm/h) was achieved, while with a high oxygen flow rate, the process was considered to be in the poisoned mode with an extremely low deposition rate (around 20 nm/h). X-ray diffraction (XRD) results show that the yttrium oxide films that were produced in the metallic mode represent a mixture of different crystal structures including the metastable monoclinic phase and the stable cubic phase, while the poisoned mode products show a dominating monoclinic phase. The thin films prepared in metallic mode have relatively dense structures with less porosity. Annealing at 600 °C for 15 h, as a structure stabilizing process, caused a phase transformation that changes the metastable monoclinic phase to stable cubic phase for both poisoned mode and metallic mode. The composition of yttrium oxide thin films changed from nonstoichiometric to stoichiometric together with a lattice parameter variation during annealing process. For the metallic mode deposition however, cracks were formed due to the thermal expansion coefficient difference between thin film and the substrate material which was not seen in poisoned mode deposition. The yttrium oxide thin films that deposited in different modes give various application options as a nuclear material.
TL;DR: In this article, the experimental results on fusion relevant pure heat load exposures of different tungsten products in the electron beam devices JUDITH 1 and 2 were summarized and a detailed postmortem analysis revealed a wide and complex range of thermally-induced surface modifications and damages, such as roughening due to plastic deformation, cracking, and melting of parts of the material surface.
Abstract: The study summarizes the experimental results on fusion relevant pure heat load exposures of different tungsten products in the electron beam devices JUDITH 1 and 2. Besides steady state heat loading, up to 106 transient ELM-like pulses were applied. A detailed postmortem analysis reveals a wide and complex range of thermally-induced surface modifications and damages, such as roughening due to plastic deformation, cracking, and melting of parts of the material surface. Different industrially available tungsten products with varying thermal and mechanical properties were investigated in order to examine their influence on the thermal shock response. Furthermore, recrystallisation of the material, which will take place during long term operation, will additionally deteriorate the mechanical strength of the plasma facing material. The results show that the mechanical strength of the material has a significant influence on the formation and evolution of damage. Especially, recrystallisation and melting/resolidification will make the material more prone to thermal shock and fatigue, accelerating the evolution of damages. The combination of different material modifications/damages accompanied by the degradation of mechanical properties will have a strong impact on the plasma performance and lifetime of plasma facing materials/components.
TL;DR: In this paper, the tungsten accumulation was suppressed in EAST tokamak to achieve a long pulse H-mode discharge by superimposing NBI-heated discharges on the NBI phase.
Abstract: EAST tokamak has been equipped with upper tungsten divertor since 2014. The tungsten accumulation has been often observed in NBI-heated H-mode discharges suggesting deleterious tungsten confinement in the plasma core. It causes not only H-L back transition but also plasma disruption in several discharges. Suppression of the tungsten accumulation is therefore the most important issue in EAST to achieve a long pulse H-mode discharge. In order to study the tungsten behavior in the long pulse discharge, tungsten spectra have been measured at 20–140 A. The tungsten density, nw, is evaluated from the intensity of tungsten unresolved transition array (W-UTA) in a wavelength range of 45–70 A which is composed of several ionization stages of tungsten, e.g. W27+-W45+ at Te0∼2.5 keV). It is found that the tungsten accumulation can be suppressed when the 4.6 GHz LHW with PLHW∼0.8 MW is superimposed on the NBI phase (PNBI = 1.9 MW). During the superimposed phase the ELM frequency, fELM, increases from ∼30 Hz to ∼60 Hz and the tungsten density is halved compared to the NBI-heated discharge. The H-mode discharge can be thus steadily sustained for longer period. It is found that the nw is a large function of the ratio of LHW power to the total injection power, PLHW/(PLHW+PNBI), and the nw can be reduced, at least, in an order of magnitude smaller than that in NBI-heated discharges at PLHW/(PLHW+PNBI) ≥ 0.8. The result strongly suggests a possible way toward the steady H-mode discharge.
TL;DR: In this paper, the results of the second experimental campaign with the ITER-Like Wall: 2013-2014 are presented for the dust survey performed at JET after the second campaign.
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.
TL;DR: In this article, it is shown that volumetric recombination and impurity radiation losses are responsible for the transition to the detached plasma regime, whereas momentum removal plays an auxiliary role providing conditions necessary for the first two to become efficient.
Abstract: Divertor plasma detachment is analyzed from the viewpoint of energy and particle balance in the edge plasma. It is shown that volumetric recombination and impurity radiation losses are responsible for the transition to the detached plasma regime, whereas “momentum removal” plays although important, but auxiliary role providing conditions necessary for the first two to become efficient. A criterion of the local (on an isolated flux tube) detachment onset is studied for both pure and impurity-seeded plasmas.
TL;DR: In this article, the ablation and fuel-retention characteristics of aluminum-tungsten (AlW) and beryllium-tooth-to-aluminum (BeW) samples have been determined using Laser Induced Breakdown Spectroscopy (LIBS) and compared to results obtained using Secondary Ion Mass Spectrometry (SIMS).
Abstract: The ablation and fuel-retention characteristics of aluminum–tungsten (AlW) and beryllium–tungsten (BeW) samples have been determined using Laser Induced Breakdown Spectroscopy (LIBS) and compared to results obtained using Secondary Ion Mass Spectrometry (SIMS). The measurements have been made both at vacuum (of the order of 10−4 Pa) and at 50 Pa of argon to especially enhance the intensities of the spectral lines of H and D. For reliable evaluation of the ablation rate of the samples the electron density ne and temperature Te of the LIBS plasma have been determined with the help of selected of spectral lines of Be, Al, and W. The electron density ne has been obtained from Stark broadening lines of Al I (308.7 nm and 394.4 nm) and Be I (457.3 nm) and Te from the Saha–Boltzmann plot using W I and W II spectral lines having a higher value of the energy of upper states in order to prevent the influence of self-absorption on the results. The results indicate similar ablation characteristics between AlW (AlWD) and BeW (BeWD) samples but the inclusion of deuterium in the coating increases the ablation rate by a factor of 10 for both sample types. Concerning fuel retention more than one order of magnitude less D is retained in the AlWD sample than in BeWD. In the presence of background argon, the H and D lines were stronger and more easily distinguishable. This is a positive sign considering the real application in ITER where LIBS measurements are foreseen to be done during maintenance breaks. However the higher pressure gave a better signal, it is still far from the measurement conditions planned for ITER which need to be tested separately.
TL;DR: In this paper, the deuterium retention is compared between pure tin and Li-Sn alloy samples in both liquid and solid states, and the results suggest that tin has a role in the retention mechanism in this material.
Abstract: The use of lithium (Li) or tin (Sn) as a liquid metal plasma facing component is proposed as a solution to the high power load issue on the divertor region of nuclear fusion reactors. The possibility to use these materials depends on their compatibility with hydrogen plasmas. With the purpose of realizing deuterium retention studies, specimens of pure Sn (99.999% Sn) and Li–Sn alloy (30 at.% Li) were exposed in the ISTTOK edge plasma. Ex situ analysis of the samples was performed by means of ion beam diagnostics. Nuclear reaction analysis (NRA) technique was applied using the D(3He,p)4He reaction to quantify the fuel retention on the samples. In this work the deuterium retention is compared between pure Sn and Li–Sn alloy samples in both liquid and solid states. All the samples were found to have retention ratios smaller than 0.1 at.%. This low retention ratio is expected for pure tin given its high mass and the instability of tin hydrides. However the retention was unexpectedly low for the case of Li–Sn which was thought to be dominated by the lithium fraction in the alloy. These results suggest that tin has a role in the retention mechanism in this material.
TL;DR: In this paper, the authors investigated the effect of divertor detachment in the TCV tokamak through experiments and modelling, and found that around the onset of detachment, the upstream density profile and outer target Dα profiles broaden, possibly leading to an increase in radiation in the SOL by increased interaction between the SOL and the carbon tiles lining the outer wall.
Abstract: Divertor detachment in the TCV tokamak has been investigated through experiments and modelling. Density ramp experiments were carried out in ohmic heated L-mode pulses with the ion ∇B drift directed away from the primary X-point, similar to previous studies [1] . Before the roll-over in the ion current to the outer strike point, C III and Dα emission from the outer leg recede slowly from the strike point toward the X-point, at a rate of ∼2.0 × 10−19 m/m−3 along the magnetic field as the electron temperature along the leg reduces with increasing density. Around the onset of detachment, the upstream density profile and outer target Dα profiles broaden, possibly leading to an increase in radiation in the SOL by increased interaction between the SOL and the carbon tiles lining the outer wall. The plasma conditions upstream and at various locations along the detached outer divertor leg have been characterised, and the consistency of this data has been checked with the interpretive OSM-EIRENE-DIVIMP suite of codes [2] and are broadly found to be consistent with measured Dγ/Dα emissivity profiles along the detached outer divertor leg.
TL;DR: The operational parameters of the divertor plasmas are insensitive to the presence of MAR, although the latter may be important for correct interpretation of the Divertor diagnostics.
Abstract: Molecule-Activated Recombination (MAR) effect is re-considered in view of divertor plasma conditions. A strong isotopic effect is demonstrated. In deuterium plasmas, the reaction chain through D2+ formation, usually considered dominant and included in 2D edge plasma models, is negligible. However, in this case the other branch, through D−, usually neglected in modelling, becomes relatively strong. The overall share of MAR in divertor plasma recycling stays within 20%. The operational parameters of the divertor plasmas, such as the peak power loading on the divertor targets or the pressure limit for partial detachment of the divertor plasma, are insensitive to the presence of MAR, although the latter may be important for correct interpretation of the divertor diagnostics.
ITER1
TL;DR: In this article, radial profiles of the ion saturation current and its fluctuation statistics are presented from probe measurements in L-mode, neutral beam heated plasmas at the outboard midplane region of KSTAR.
Abstract: Radial profiles of the ion saturation current and its fluctuation statistics are presented from probe measurements in L-mode, neutral beam heated plasmas at the outboard mid-plane region of KSTAR. The results are consistent with the familiar two-layer structure, seen elsewhere in tokamak L-mode discharges, with a steep near-SOL profile and a broad far-SOL profile. The profile scale length in the far-SOL increases drastically with line-averaged density, thereby enhancing plasma interactions with the main chamber walls. Time series from the far-SOL region are characterised by large-amplitude bursts attributed to the radial motion of blob-like plasma filaments. Analysis of a data time series of several seconds duration under stationary plasma conditions reveals the statistical properties of these fluctuations, including the rate of level crossings and the average duration of periods spent above a given threshold level. This is shown to be in excellent agreement with predictions of a stochastic model, giving novel predictions of plasma–wall interactions due to transient transport events.
TL;DR: In this article, a liquid metal infused capillary porous structures (CPSs) are considered as a potential divertor solution for DEMO due to their potential power handling capability and resilience to long term damage.
Abstract: Liquid metal infused capillary porous structures (CPSs) are considered as a potential divertor solution for DEMO due to their potential power handling capability and resilience to long term damage. In this work the power handling and performance of such Sn-based CPS systems is assessed both experimentally and via modelling. A Sn-CPS target was exposed to heat fluxes of up to 18.1 MW m−2 in He plasma in the Pilot-PSI linear device. Post-mortem the target showed no damage to nor any surface exposure of the underlying W-CPS felt. The small pore size (∼40 µm) employed resulted in no droplet formation from the target in agreement with calculated Rayleigh-Taylor and Kelvin-Helmoholtz instability thresholds. The temperature response of the Sn-target was used to determine the thermal conductivity of the mixed Sn-CPS material using COMSOL modelling. These values were then used via further finite element analysis to extrapolate to DEMO relevant monoblock designs and estimate the maximum power handling achievable based on estimated temperature windows for all component elements of the design. For an optimized design a heat-load of up to 20 MW m−2 may be received while the use of CPS also offers other potential design advantages such as the removal of interlayer requirements.
TL;DR: In this paper, a 2D UEDGE transport model for DIII-D tokamak divertor plasmas is presented, which includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode.
Abstract: A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode. The data set, which spans a range of plasma densities for both forward and reverse toroidal magnetic field (Bt), is provided by divertor Thomson scattering (DTS). Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (Te) and density (ne) across both divertor legs for individual discharges. The simulations focus on the open magnetic field-line regions, though they also include a small region of closed field lines. The calculations show the same features of in/out divertor plasma asymmetries as measured in the experiment, with the normal Bt direction (ion ∇B drift toward the X-point) having higher ne and lower Te in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. These 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.
TL;DR: The Swiss Plasma Center (SPC) is planning a divertor upgrade for the TCV tokamak as discussed by the authors, which aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance.
Abstract: The Swiss Plasma Center (SPC) is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.
TL;DR: In this paper, a Divertor Spectroscopy System has been developed for the purpose of analysis of the Balmer line emission at TCV with emphasis on analysis of Balmer 6 → 2 and 7 → 2 line emission.
Abstract: The aim of this work is to provide an understanding of detachment at TCV with emphasis on analysis of the Balmer line emission. A new Divertor Spectroscopy System has been developed for this purpose. Further development of Balmer line analysis techniques has allowed detailed information to be extracted from the three-body recombination contribution to the n = 7 Balmer line intensity. During density ramps, the plasma at the target detaches as inferred from a drop in ion current to the target. At the same time the Balmer 6 → 2 and 7 → 2 line emission near the target is dominated by recombination. As the core density increases further, the density and recombination rate are rising all along the outer leg to the x-point while remaining highest at the target. Even at the highest core densities accessed (Greenwald fraction 0.7) the peaks in recombination and density may have moved not more than a few cm poloidally away from the target which is different to other, higher density tokamaks, where both the peak in recombination and density continue to move towards the x-point as the core density is increased. The inferred magnitude of recombination is small compared to the target ion current at the time detachment (particle flux drop) starts at the target. However, recombination may be having more localized effects (to a flux tube) which we cannot discern at this time. Later, at the highest densities achieved, the total recombination does reach levels similar to the particle flux.
TL;DR: In this article, post-mortem analyses of individual components provide relevant information on plasma-surface interactions like tungsten erosion, beryllium deposition and plasma fuel retention with divertor tiles.
Abstract: Post-mortem analyses of individual components provide relevant information on plasma-surface interactions like tungsten erosion, beryllium deposition and plasma fuel retention with divertor tiles v ...
TL;DR: In this paper, the deuterium retention was measured in tungsten samples simultaneously irradiated by W ions and exposed to D atoms at five different temperatures from 450 to 1000 k. The results were compared to different sequential damaging/exposure experiments.
Abstract: Deuterium retention was for the first time measured in tungsten samples simultaneously irradiated by W ions and exposed to D atoms at five different temperatures from 450 K to 1000 K. In order to obtain information on the defect concentration, samples were afterwards exposed to D atoms at 600 K to populate the created defects. The results were compared to different sequential damaging/exposure experiments. Synergistic effects were observed, namely, higher D concentrations were found in the case of simultaneous damaging and D-atom loading as compared to sequential damaging at elevated temperatures and populating the defects afterwards. However, the deuterium retention is still lower as compared to sequential damaging at room temperature and post-damaging annealing. The observations are explained by stabilization of defects by the presence of solute hydrogen in the bulk that would annihilate at high temperatures without the presence of hydrogen. Results of simultaneous W-ion damaging and D exposure at elevated temperatures were also compared to a sequential experiment of W-ion damaging at room temperature and then D-atom loading at high temperatures showing that thermal D de-trapping dominates deuterium retention at high temperatures.
TL;DR: In this paper, the authors used ion orbit modelling to calculate the heating of tungsten monoblocks during ELMs at the inner vertical target, where the highest surface energy densities are expected.
Abstract: The high heat flux areas on the vertical divertor targets in the ITER tokamak will consist of cuboid tungsten monoblocks bonded to copper cooling tubes Three-dimensional ion orbit modelling is used to calculate the heating of tungsten monoblocks during ELMs at the inner vertical target, where the highest surface energy densities are expected The presence of thin gaps between monoblocks results in exposed edges onto which the heat flux can be focused ELM ions are focused by their gyromotion onto the magnetically shadowed, long toroidal edges of the monoblocks The risk of monoblock edge melting is greater than the risk of full surface melting on the plasma-wetted zone Alternative shaping solutions such as edge chamfering, filleting, and poloidal beveling do not show promise; the melt zone simply migrates to other locations on the monoblocks Without ELM mitigation, there is a marginal risk of edge melting due to uncontrolled ELMs in the pre-nuclear phase of ITER operation, and an absolute certainty of it in the burning nuclear phase To avoid edge melting altogether, the surface energy density would have to limited to less than 015 MJ/m2
TL;DR: In this paper, a theoretical description of the tungsten erosion at disruption in tokamak plasma is presented, and the proposed model was verified by comparison with experimental observations in the T-10.
Abstract: Full tungsten poloidal and mushroom limiters were tested in series of experiments with disruptions in the T-10 tokamak. Significant melting, formation of small craters and erosion of the tungsten limiter have been observed after ∼400 discharges with disruption. A theoretical description of the tungsten erosion at disruption in tokamak plasma is presented. The proposed model was verified by comparison with experimental observations in the T-10. The results are used for the erosion prediction of the ITER tungsten divertor.
TL;DR: In this article, the European Fusion Roadmap foresees water cooled plasma facing components in a first DEMO design in order to provide enough margin for the cooling capacity and to only moderately extrapolate the technology which was developed and tested for ITER.
Abstract: The European Fusion Roadmap foresees water cooled plasma facing components in a first DEMO design in order to provide enough margin for the cooling capacity and to only moderately extrapolate the technology which was developed and tested for ITER. In order to make best use of the water cooling concept copper (Cu) and copper-chromium-zirconium alloy (CuCrZr) are envisaged as heat sink whereas as armour tungsten (W) based materials will be used. Combining both materials in a high heat flux component asks for an increase of their operational range towards higher temperature in case of Cu/CuCrZr and lower temperatures for W. A remedy for both issues- brittleness of W and degrading strength of CuCrZr- could be the use of W fibres (Wf) in W and Cu based composites. Fibre preforms could be manufactured with industrially viable textile techniques. Flat textiles with a combination of 150/70 µm W wires have been chosen for layered deposition of tungsten-fibre reinforced tungsten (Wf/W) samples and tubular multi-layered braidings with W wire thickness of 50 µm were produced as a preform for tungsten-fibre reinforced copper (Wf /Cu) tubes. Cu melt infiltration was performed together with an industrial partner resulting in sample tubes without any blowholes. Property estimation by mean field homogenisation predicts strongly enhanced strength of the Wf/CuCrZr composite compared to its pure CuCrZr counterpart. Wf /W composites show very high toughness and damage tolerance even at room temperature. Cyclic load tests reveal that the extrinsic toughening mechanisms counteracting the crack growth are active and stable. FEM simulations of the Wf/W composite suggest that the influence of fibre debonding, which is an integral part of the toughening mechanisms, and reduced thermal conductivity of the fibre due to the necessary interlayers do not strongly influence the thermal properties of future components.