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Abstract: Abstract The Vacuum Vessel (VV), as the first confinement barrier of the Chinese Fusion Engineering Test Reactor (CFETR) must guarantee its integrity once the In Vacuum Vessel Loss of Coolant Accident (In-VV LOCA) happens. Enough experimental verifications are inevitable. And the data from the experimental facility should be convincing with an acceptable cost. In the present research, scaling criteria are identified for the In-VV LOCA, which can provide the direction for the design scheme of the experimental facility. The experimental facility employs the same working fluid and the equal thermal operation conditions. The volume scaling factor of 1/1000 for the VV with a toroidal structure adopted in the experimental facility is selected. The similarities of the flow at the break, flash and the flow resistance are prioritized. Through the computational simulation for the experimental facility, it has been demonstrated that the thermal hydraulic responses could be reappeared in the experimental facility within an acceptable range of distortions. PubDate: 2025-02-15 DOI: 10.1007/s10894-025-00480-9
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Abstract: Abstract The lithium vapor divertor concept is being developed as a method to achieve detached divertor conditions in a tokamak while minimizing impurity radiation losses from the core plasma. SOLPS-ITER modeling has previously been used to identify some of the geometric constraints and required lithium evaporation rate of a lithium vapor divertor in a medium-sized tokamak during steady-state operation. Here an updated conceptual design based on these operating requirements is introduced and the thermal response of the system is modeled during cyclical operation, consistent with operation in a short-pulse tokamak. Controllability of the temperature of the lithium capillary porous system (CPS) is achieved by adopting a design where there is no line-of-sight for radiation from the plasma to reach the heated CPS surface. Operational strategies to minimize the amount of lithium evaporated between plasma discharges while achieving steady evaporation rates during plasma discharges are discussed and modeled here. The optimal feedforward control strategy demonstrated in this work is to ramp up the temperature of the evaporator as quickly as possible immediately before a plasma discharge and then reduce the heating to match the desired steady-state net evaporation rate just before the plasma discharge begins, allowing the thermal inertia of the system to stabilize the evaporation rate during the first second of the plasma discharge. PubDate: 2025-02-14 DOI: 10.1007/s10894-025-00479-2
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Abstract: Abstract 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 (Bt) 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 Bt, 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 Bt. In favourable Bt 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 Zeff inside the separatrix and has an intermediate flow rate for detachment at the outer target, comparing with the X-point and upstream locations. PubDate: 2025-02-10 DOI: 10.1007/s10894-025-00477-4
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Abstract: Abstract 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 Ti ~ 200 keV, electron temperatures Te ~ 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. PubDate: 2025-02-04 DOI: 10.1007/s10894-025-00478-3
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Abstract: Abstract 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. PubDate: 2025-02-01 DOI: 10.1007/s10894-025-00476-5
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Abstract: Abstract 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. PubDate: 2025-01-22 DOI: 10.1007/s10894-025-00474-7
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Abstract: Abstract 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. PubDate: 2024-12-10 DOI: 10.1007/s10894-024-00472-1
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Abstract: Abstract 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. PubDate: 2024-12-10 DOI: 10.1007/s10894-024-00473-0
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Abstract: Abstract 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. PubDate: 2024-10-29 DOI: 10.1007/s10894-024-00465-0
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Abstract: Abstract The Superconducting Conductor Testing Facility, which is developed to evaluate the reliability of engineering technology and safe operation in a fusion reactor operation environment, is under construction by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). Since the Nb3Sn layer coil of this test facility adopts the manufacturing process of the wind & react, the high-strength glass fiber used as the inter-turn insulation material will carbonize after high-temperature heat treatment at 665℃, thereby reducing the mechanical and electrical properties of the winding. The surface decarburization and modification process of high-strength glass fiber was developed to improve the properties of glass fiber after heat treatment. It is verified that the developed glass fiber tape treatment process can meet manufacturing process requirements of layered Nb3Sn superconducting magnets through the coil winding radial pressure test and VPI sample performance test. This processing technology has been successfully applied in the manufacturing of experimental magnets, providing technical support for the insulation manufacturing of a large Nb3Sn layer coil. PubDate: 2024-10-29 DOI: 10.1007/s10894-024-00471-2
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Abstract: Abstract Radiative heat transfer is a fundamental process in high energy density physics and inertial fusion. Accurately predicting the behavior of Marshak waves across a wide range of material properties and drive conditions is crucial for design and analysis of these systems. Conventional numerical solvers and analytical approximations often face challenges in terms of accuracy and computational efficiency. In this work, we propose a novel approach to model Marshak waves using Fourier Neural Operators (FNO). We develop two FNO-based models: (1) a base model that learns the mapping between the drive condition and material properties to a solution approximation based on the widely used analytic model by Hammer & Rosen (2003), and (2) a model that corrects the inaccuracies of the analytic approximation by learning the mapping to a more accurate numerical solution. Our results demonstrate the strong generalization capabilities of the FNOs and show significant improvements in prediction accuracy compared to the base analytic model. PubDate: 2024-10-24 DOI: 10.1007/s10894-024-00470-3
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Please help us test our new pre-print finding feature by giving the pre-print link a rating. A 5 star rating indicates the linked pre-print has the exact same content as the published article.
Please help us test our new pre-print finding feature by giving the pre-print link a rating. A 5 star rating indicates the linked pre-print has the exact same content as the published article.
Please help us test our new pre-print finding feature by giving the pre-print link a rating. A 5 star rating indicates the linked pre-print has the exact same content as the published article.
Please help us test our new pre-print finding feature by giving the pre-print link a rating. A 5 star rating indicates the linked pre-print has the exact same content as the published article.
Abstract: Abstract Open field line currents are intrinsic to DC helicity injection plasma startup and pose a challenge for inferring the plasma equilibrium with standard reconstruction analysis. Local helicity injection (LHI) is a type of DC helicity injection which uses small, modular current sources to drive force-free current along helical field lines to produce tokamak plasmas. MHD modeling and magnetic measurements during LHI indicate the injected current streams remain coherent as helical structures on the outboard edge of a core toroidal plasma that is tokamak-like in a toroidally averaged sense. To extract core plasma equilibrium properties, external magnetic diagnostics corrected for contributions from the injected current streams are fitted by a standard Grad-Shafranov equilibrium code. An iterative approach for estimating and subtracting the stream contributions from the diagnostic signals is described and applied to a model equilibrium database to reduce systematic errors introduced by the streams. Convergence is usually attained with 2 to 4 iterations, with derived equilibrium parameters matching the prescribed axisymmetric core values to within estimated experimental uncertainties. Accurate recovery of core parameters occurs when the ratio of the net toroidal windup current from the streams to the core plasma current is less than 0.2, which is typically satisfied in most experiments. PubDate: 2024-10-12 DOI: 10.1007/s10894-024-00460-5
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