Journal of Fusion Energy最新文献

The effects of B powder injection on plasma detachment about EAST discharge were studied by using SOLPS-ITER code package with the effects of E × B drifts considered. The simulation results show that plasma detachment occurs at the inner target in favourable toroidal magnetic field (B_t) direction at a relatively low B powder flow rate, one order of magnitude lower than that at the outer target. In a similar scenario with unfavourable B_t, it is found that the detachment thresholds of B flow rate for both the inner and outer targets are close and of the same order as that for the outer target with favourable B_t. In favourable B_t direction at B powder flow rate of 1.2 × 10²¹ atoms/s, a localized, broadened high-density region is formed near the inner target benefitted by the injection location and the E × B drift, and a radiation-intensified zone, mostly contributed by B¹⁺ and B²⁺, occurs there. The E × B drift facilitates plasma detachment at the inner target and simultaneously amplifies the in–out divertor asymmetry. In addition, the simulation results with three different injection locations show that the injection from outer strike point leads to the lowest Z_eff inside the separatrix and has an intermediate flow rate for detachment at the outer target, comparing with the X-point and upstream locations.

The features of high-gain aneutronic p–¹¹B fusion are examined. A comparison of inertial systems with extremely high plasma densities (n ~ 10³⁰–10³¹ m^–3) and stationary systems with magnetic confinement of low-density plasma (n ~ 10²⁰–10²² m^–3) shows that it is also necessary to analyze combined schemes based on magneto-inertial systems with fuel refill. Present work considers limiting modes of the quasi-stationary phase of fusion, which show the maximum plasma gain at plasma density n ~ 10³¹ m^–3 and ion temperatures T_i ~ 200 keV, electron temperatures T_e ~ 100 keV at the beginning of the quasi-stationary phase. The content of reaction products (α-particles) has a significant influence on the parameters of the system. If the confinement time of α-particles is the same as for the fuel components, then due to radiation the energy gain Q ~ 1. In modes with a reduced confinement time of α-particles, the gain reaches a value of Q ~ 6. A further increase in Q requires extremely high plasma energy.

Liquid metals, like lithium (Li), are considered a promising plasma-facing material due to their self-repairing, in comparison with the conventional solid materials that have limitations in handling high heat flux in future fusion devices. To predictively simulate global Li transport under the lithium divertor condition, the three-dimensional Monte Carlo code ITCD has been upgraded significantly, in terms of the simulation domain (from the sole divertor region in a limited toroidal range to the entire edge plasma region in a full toroidal torus). The expansion of the simulation zone brings about the new demand of the computational resource, which motivates us to implement the guiding-center (GC) particle push approach into ITCD. The trajectory of charged Li particle using the GC particle push approach shows a good agreement with the full-orbit (FO) particle push method. The FO and GC hybrid particle push scheme has been used to deal with the gyration scrape-off effect and meanwhile speed up the calculation of the global Li transport. The characteristics of Li impurity density and deposition distributions are studied in detail by ITCD.

First experiments are reported of the simultaneous exposure of a number of Sn-wetted W CPSs and a reference W CPS to 100 ms NBI pulses (divertor steady-state loading conditions) and 2 ms long high-energy laser pulses (divertor ELM like loading conditions) at the High-Heat Flux OLMAT facility. The use of a fast-frame imaging camera allows monitoring the onset of particle ejection from the targets during laser pulses and obtaining the corresponding laser heat fluxes as a measure of the resilience of these targets. Fast camera images are used also to determine ejected particle numbers and to estimate their maximum velocities as laser power is increased in order to compare the influence of W CPS structure on these parameters. In addition, the craters resulting from particle ejection are studied for each target with an optical microscope and a scanning electron microscope. Moreover, in-situ W and Sn particle ejection is followed using visible emission spectroscopy and post-exposure W melting after particle ejection is observed using the energy dispersive X-ray method EDX for all the studied targets. This shows that Sn is unable to protect the underlying W substrate from high-energy laser damage, albeit a subsequent refilling of the formed craters with Sn is visible during NBI-only pulses after laser damage. Thus, it is considered that optimization of surface refilling/replenishment with Sn is needed to improve the W substrate protection. From this work, it is also found that the W CPS reference material has a higher laser heat flux threshold for particle ejection than the Sn-wetted targets. Nevertheless, it is important to take into account that in these experiments with laser pulses, the possible beneficial effects of vapor shielding that can take place during particle irradiation at ELMs or disruptions are not present, thus these experiments represent a worst-case scenario.

Oxidation resistant smart tungsten alloys (SA) are considered a promising plasma facing material for DEMO reactors. SA-based plasma facing components (PFC) have to meet several long-term operation requirements. Among other criteria, these PFC should be able to withstand high thermal loads and be corrosion resistant in liquid lithium for a liquid first wall design implementation. In this work, smart tungsten alloys WCrY, WCrZr were brazed to reduced activation ferritic-martensitic (RAFM) steels Eurofer97, CLAM via 48Ti–48Zr–4Be wt.% brazing alloy. Thermal stability of the brazed joints was investigated. High temperature shear tests at 300, 600 °C were carried out. The shear strength of WCrZr/Ta/CLAM joints is 50 ± 4 and 49 ± 5 MPa at 300 and 600 °C, respectively. Unbrazing of the WCrY/Ta/Eurofer97 and WCrZr/Ta/CLAM joints occurs at 1447 and 1522 °C, respectively, due to the melting of steels. Corrosion resistance of the smart tungsten alloys, SA/Ta/RAFM joints in liquid lithium at 600 °C, 100 h exposure was demonstrated.

Vertical position control of tokamak plasmas is essential for exploring operational limits and ensuring stable operation at high elongations to avoid disruptions. This study focuses on improving vertical instability control in the HL-3 tokamak by enhancing the signal-to-noise ratio of control signals and optimizing control strategies. We employed improved diagnostic techniques using Mirnov coils and flux loops, combined with digital filtering technology, to mitigate the effects of power supply switching and measurement noise. The vertical stabilization (VS) control system was upgraded with an optimized low-pass filter for vertical position estimation, a novel method for vertical velocity estimation using direct voltage signals from diagnostics, and an improved control algorithm. These enhancements resulted in significant improvements in control precision and noise reduction. Experimental results demonstrated successful control of highly elongated plasmas ((kappa ) up to 1.8) with high plasma currents (up to 1.6 MA), achieving vertical position control accuracy better than 1 cm during the plasma current ramp-up phase. These advancements expand the operational parameter space of HL-3, paving the way for higher performance plasma operation.

The ITER Electron Cyclotron Heating and Current Drive (ECH) plays a pivotal role in heating and controlling fusion plasmas, with the Steering Mirrors being a crucial component of this actuator. A representative model of the ECH is compulsory in the development and validation of the Plasma Control System (PCS). This manuscript aims to propose a Control-Oriented model of the Steering Mirrors based on the design tested at the Swiss Plasma Centre. In this design a steering mirror rotates on some frictionless flexure pivots due to the action of a set of externally pressurized bellows acting against pre-compressed springs. This system is referred to as the Steering Mirror Assembly (SMA). The adherence of the model is tested by comparing the simulations with the experimental results, while considering ITER’s most recent requirements. Performances, generally increased in terms of accuracy, are in line with the prototype’s results.