ELM-induced transient tungsten melting in the JET divertor
TL;DR: In this paper, the outer strike point of a full-tungsten (W) divertor was moved towards the leading edge of the W divertor and the base temperature was raised within 1 s to a level allowing transient melting during the subsequent 0.5 s.
Abstract: The original goals of the JET ITER-like wall included the study of the impact of an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek et al 2013 Nucl. Fusion 53 083023). ITER has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key inputs required in support of this decision was the study of the possibility of W melting and melt splashing during transients. Damage of this type can lead to modifications of surface topology which could lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. JET is able to produce ELMs large enough to allow access to transient melting in a regime of relevance to ITER. Transient W melt experiments were performed in JET using a dedicated divertor module and a sequence of I-P = 3.0 MA/B-T = 2.9 T H-mode pulses with an input power of P-IN = 23 MW, a stored energy of similar to 6 MJ and regular type I ELMs at Delta W-ELM = 0.3 MJ and f(ELM) similar to 30 Hz. By moving the outer strike point onto a dedicated leading edge in the W divertor the base temperature was raised within similar to 1 s to a level allowing transient, ELM-driven melting during the subsequent 0.5 s. Such ELMs (delta W similar to 300 kJ per ELM) are comparable to mitigated ELMs expected in ITER (Pitts et al 2011 J. Nucl. Mater. 415 (Suppl.) S957-64). Although significant material losses in terms of ejections into the plasma were not observed, there is indirect evidence that some small droplets (similar to 80 mu m) were released. Almost 1 mm (similar to 6 mm(3)) of W was moved by similar to 150 ELMs within 7 subsequent discharges. The impact on the main plasma parameters was minor and no disruptions occurred. The W-melt gradually moved along the leading edge towards the high-field side, driven by j x B forces. The evaporation rate determined from spectroscopy is 100 times less than expected from steady state melting and is thus consistent only with transient melting during the individual ELMs. Analysis of IR data and spectroscopy together with modelling using the MEMOS code Bazylev et al 2009 J. Nucl. Mater. 390-391 810-13 point to transient melting as the main process. 3D MEMOS simulations on the consequences of multiple ELMs on damage of tungsten castellated armour have been performed. These experiments provide the first experimental evidence for the absence of significant melt splashing at transient events resembling mitigated ELMs on ITER and establish a key experimental benchmark for the MEMOS code.
Citations
More filters
TL;DR: In this paper, the main focus is on the main design driver, steady state power fluxes in the DT phases, obtained from simulations using the 2-D SOLPS-4.3 and SolPS-ITER plasma boundary codes, assuming the use of the low Z seeding impurities nitrogen (N) and neon (Ne).
Abstract: On the eve of component procurement, this paper discusses the present physics basis for the first ITER tungsten (W) divertor, beginning with a reminder of the key elements defining the overall design, and outlining relevant aspects of the Research Plan accompanying the new “staged approach” to ITER nuclear operations which fixes the overall divertor lifetime constraint. The principal focus is on the main design driver, steady state power fluxes in the DT phases, obtained from simulations using the 2-D SOLPS-4.3 and SOLPS-ITER plasma boundary codes, assuming the use of the low Z seeding impurities nitrogen (N) and neon (Ne). A new perspective on the simulation database is adopted, concentrating purely on the divertor physics aspects rather than on the core-edge integration, which has been studied extensively in the course of the divertor design evolution and is published elsewhere. Emphasis is placed on factors which may increase the peak steady state loads: divertor target shaping for component misalignment protection, the influence of fluid drifts, and the consequences of narrow scrape-off layer heat flux channels. All tend to push the divertor into an operating space at higher sub-divertor neutral pressure in order to remain at power flux densities acceptable for the target material. However, a revised criterion for the maximum tolerable loads based on avoidance of W recrystallization, sets an upper limit potentially ∼50% higher than the previously accepted value of ∼10 MW m−2, a consequence both of the choice of material and the finalized component design. Although the simulation database is currently restricted to the 2-D toroidally symmetric situation, considerable progress is now also being made using the EMC3-Eirene 3-D code suite for the assessment of power loading in the presence of magnetic perturbations for ELM control. Some new results for low input power corresponding to the early H-mode operation phases are reported, showing that even if realistic plasma screening is taken into account, significant asymmetric divertor heat fluxes may arise far from the unperturbed strike point. The issue of tolerable limits for transient heat pulses is an open and key question. A new scaling for ELM power deposition has shown that whilst there may be more latitude for operation at higher current without ELM control, the ultimate limit is likely to be set more by material fatigue under large numbers of sub-threshold melting events.
312 citations
TL;DR: In this paper, the authors reviewed the 2014-2016 JET results in the light of their significance for optimising the ITER research plan for the active and non-active operation, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric.
Abstract: The 2014-2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L-H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H = 1 at β N ∼ 1.8 and n/n GW ∼ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D-T campaign and 14 MeV neutron calibration strategy are reviewed.
162 citations
Cites background from "ELM-induced transient tungsten melt..."
...For this first set of experiments both the local thermal response and the observed melt motion could only be explained under the assumption of a significantly (60–80%) lower heat flux to the exposed leading edge than expected from purely geometrical projection of the parallel heat flux derived from thermography data at standard tile surfaces [58]....
[...]
140 citations
134 citations
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
References
More filters
TL;DR: For 47 elements in the range 2 ≤ Z ≤ 92, steady-state radiative cooling rates, average charge states, and mean square charge states have been calculated for low-density, high-temperature plasmas (n e ≲ 10 16 electrons/cm 3 and T = 0.002-100 keV) as discussed by the authors.
743 citations
TL;DR: In this article, the authors consider the risks engendered by the baseline divertor strategy with regard to known W plasma-material interaction issues and briefly present the current status of a possible full-tungsten (W) divertor design.
610 citations
TL;DR: The Sustained Spheromak Physics Experiment (SSPX) as discussed by the authors was a high-temperature (Te up to 0.5 keV), coaxial helicity injection (CHI) formed by coaxial helical injection, with plasma duration of a few milliseconds following the high-current formation stage.
Abstract: The Sustained Spheromak Physics Experiment (SSPX) was a high-temperature (Te up to 0.5 keV) spheromak formed by coaxial helicity injection (CHI) and with plasma duration of a few milliseconds following the high-current formation stage. Clean walls and low impurity operation were obtained by a combination of baking, discharge cleaning and titanium deposition on the walls, allowing the generation of high-quality plasmas. Resistive-magnetohydrodynamic simulations, benchmarked to the experiment, were used to elucidate the physics. The detailed characteristics of the nφ = 1 toroidal mode associated with CHI were determined as was the physics of the nonlinear current drive and magnetic reconnection that formed and sustained the spheromak. If the helicity injection rate was reduced following formation the plasma became relatively quiescent and magnetic surfaces formed. The measured thermal diffusivity in the core was as low as ∼1 m2 s−1. However, reconnection events during buildup or sustainment of the plasma current by CHI were found to open magnetic surfaces throughout the plasma allowing rapid energy loss to the walls. As a result, experiments and simulations in SSPX found no path to simultaneous sustainment by CHI and good energy confinement. Additional physics results are also presented in this review.
482 citations
TL;DR: The ITER PFC design has now reached a rather mature stage following the 2007 ITER Design Review as discussed by the authors, and the key elements of the design, reviews the physics drivers, essentially thermal load specifications, which have defined the concept and discusses a selection of material and design issues.
368 citations
TL;DR: In this article, the ITER-like wall (ILW) experiment at JET was used to demonstrate the plasma compatibility with metallic walls and the reduction in fuel retention, which confirmed the expected predictions concerning the plasma-facing material change in ITER and is in line with identification of fuel co-deposition with Be as the main mechanism for the residual long-term retention.
Abstract: JET underwent a transformation from a full carbon-dominated tokamak to a fully metallic device with beryllium in the main chamber and a tungsten divertor. This material combination is foreseen for the activated phase of ITER. The ITER-Like Wall (ILW) experiment at JET shall demonstrate the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments (Ip = 2.0 MA, Bt = 2.0–2.4 T, δ = 0.2–0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction in the long-term retention rate by more than a factor of 10 with respect to carbon-wall reference discharges. All experiments are executed in a series of identical plasma discharges in order to achieve maximum plasma duration until the analysis limit of the active gas handling system is reached. The composition analysis shows high purity of the recovered gas, typically 99% D. For typical L-mode discharges (Paux = 0.5 MW), type III (Paux = 5.0 MW) and type-I ELMy H-mode plasmas (Paux = 12.0 MW) a drop of the deuterium retention rate normalized to the operational time in divertor configuration is measured from 1.27 × 1021, 1.37 × 1021 and 1.97 × 1021 D s−1 down to 4.8 × 1019, 7.2 × 1019 and 16 × 1019 D s−1, respectively. The dynamic retention increases in the limiter phase in comparison with carbon-fibre composite, but also the outgassing after the discharge has risen in the same manner and overcompensates this transient retention. Overall an upper limit of the long-term retention rate of 1.5 × 1020 D s−1 is obtained with the ILW. The observed reduction by one order of magnitude confirms the expected predictions concerning the plasma-facing material change in ITER and is in line with identification of fuel co-deposition with Be as the main mechanism for the residual long-term retention. The reduction widens the operational space without active cleaning in the DT phase in comparison with a full carbon device.
194 citations