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Abstract: Abstract The liquid metal magnetohydrodynamic (MHD) flow through coupled ducts with conducting walls under inclined transversal gradient magnetic field is an important physical flow phenomenon, which has the unknown physical mechanism about the interaction between the electromagnetic coupling effect and the three-dimensional (3D) MHD effect. To reveal this physical mechanism, 3D numerical simulations based on a customized solver in the OpenFoam environment are conducted to systematically study the effect of inclined gradient magnetic field on the MHD flow states through coupled conducting ducts. Then the mechanism behind the generation of the 3D MHD effect in the gradient magnetic field zones has been discussed in detail. It is found that the electromagnetic coupling effect can enhance this 3D MHD effect in the co-flow case, but suppress it in the counter-flow case. Moreover, the strong electromagnetic coupling effect in the counter-flow case will induce a “self-circulation” flow region in the duct when the external magnetic field is inclined, and the inclined angle also has a great influence on the area of this flow region, which reduces with the increase of the inclined angle. These results are important for the in-depth fundamental understanding of the 3D MHD effect of liquid metal flowing through coupled conducting ducts under inclined gradient magnetic field, and also helpful for the future design of the liquid metal blanket of fusion reactor. PubDate: 2023-11-27
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Abstract: Abstract Liquid metals corrode structure materials in fusion, fission, and spallation applications. A duct strongly cooled on the outside surface is proposed to mitigate or eliminate the corrosion problem. A solidified metal layer between the cool duct (Tduct<Tmelt) and the liquid metal could serve as an interface to protect the duct from corrosion. PubDate: 2023-11-14
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Abstract: Abstract When the local heat flux exceeds specified flux limit, tungsten PFC surfaces can be damaged, which is not acceptable for a reliable reactor operations. The divertor PFCs are typically designed for a specific heat flux limit usually assuming an average steady-state heat flux which is typically 5–10 MW/m2. However, in addition to steady-state heat flux, fusion reactor divertor PFCs could experience transient heat fluxes such as ELMs and/or other magnetic reconnection events which can deposit large transient heat fluxes onto the divertor PFCs. The transient divertor heat flux could be significantly larger than the steady-state heat flux which could damage the solid PFC surfaces. The divertor heat flux can be subjected to additional complications such as the uncertainties in the the divertor strike point heat flux projection. Moreover, there are additional experimental observations of non-axisymmetric power flux which can occur under non-axisymmetric magnetic perturbations. The liquid lithium (LL) PFCs is more resilient against such transient heat fluxes as they could evaporate LL as needed and the lost LL can be then replenished afterward. In this paper, we analyze a case for a transient divertor heat pulse of 1 MJ in 10 ms for a ITER-size reactor. This is a small perturbation (~ 0.1%) to the expected plasma stored energy compared to the previously analyzed case of 20 MJ heat pulse. Even with this relatively modest heat pulse, the LL surface undergoes ~ 100 °C temperature rise. However, the resulting LL surface heating without rapid cooldown mechanism could lead to excessive LL evaporation continuing well after the transient heat flux resulting in a significant Li injection of ~ 0.6 mol in about a 200 ms period. This amount of Li injection could cause plasma dilution and performance degradation. On the other hand, an active Li injection capability if optimized could prevent the LL surface temperature rise and thus reducing subsequent Li evaporation into the plasma by a factor of 7 compared to the passive LL PFC case. A crucial tool of active Li injection is a rapid response pellet injector which could inject light impurity pellets before the excessive heat flux could reach the divertor plate causing serious damage. A simple pellet ablation model suggests a favorable pellet deposition profile for smaller ~ 0.1 mm radius pellet with ~ 10–20 m/s velocity. Moreover, if it is possible to inject from the private flux region, the pellet injection efficiency into the high heat flux strike point region can be as high as 80% compared to ~ 50% for the injection from outer radius region. The pellet deposition efficiency can be further improved by designing a shell-pellet which can burst when a certain ablation fraction is reached. A possible implementation technique using an inductive pellet injector with a rapid time response of a few msec is proposed here which can be tested in NSTX-U. PubDate: 2023-10-12
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Abstract: Abstract Harsh heat load conditions on plasma-facing components (PFCs) in steady-state and transient phenomena (e.g., disruptions and ELMs) in DEMO fusion reactors question the feasibility of current approaches based on solid targets made of tungsten. This issue calls for the development of innovative plasma-facing components. Liquid metal PFCs with strong convection enhance heat removal capability and resilience after the transient phenomena. However, transporting liquid metal across magnetic fields gives rise to MHD drag. MHD drag for the case of uniform B, estimated analytically, is acceptable. Grad-B MHD drags with straight ducts could seriously drag the LM flow across non-uniform B. Expanding the duct along B and shrinking the duct in a perpendicular direction could make electromotive force vBh approximately constant along the duct and significantly reduces the grad-B MHD drag. Here v denotes the flow velocity along the duct, B is the magnetic field strength, and h is the vertical duct size. Three-dimensional simulations for internal and free surface thermo-MHD phenomena have demonstrated that the proposed duct design reduces the total pressure drop along the duct. PubDate: 2023-10-10
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Abstract: Abstract A new control system of shattered pellet injection (SPI) has been successfully developed and implemented in the experimental advanced superconducting tokamak. The control system comprises four functional modules responsible for vacuum acquisition, temperature regulation, gas supply, and system protection, which facilitate the safe and stable operation of the SPI. The software framework employed for the SPI control system incorporates experimental physics and industrial control system and Phoebus. Utilizing these integrated control systems, the gun barrel temperature and material gas pressure could be accurately controlled during pellet forming phase. Also, it could cooperatively control the various types of valves to achieve material gas supply, propellant gas supply and timely pumping. Finally, the pellet was successfully generated, separated from the gun barrel, and accelerated into the plasma vacuum vessel controlled by this system. In addition, the issue of extended delay time was observed in SPI experiments, and a potential solution is also proposed in the paper. PubDate: 2023-10-07
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Abstract: Abstract When the breakdown of acceleration grid occurs, the core snubber can consume fault energy and limit the arc current. In this paper, the expression of equivalent circuit model is optimized. The value of magnetization inductance at da1/dt = 0 is analyzed and the equivalent resistance expression eddy current coefficient is revised from 2.5 to 1. The simulation model with gyrator and the high voltage short circuit experiment platform are described. The rationality of the model hypothesis is verified. Then the effectiveness of equivalent circuit model established in this paper is verified. Also, the dynamic response of the discharge circuit is obtained and the influence of system leakage inductance on the surge suppression process is analyzed. At the same time, in this paper, the distributed capacitance of the transmission line is extracted, and the structure parameters of core snubber suitable for −1MV voltage level are designed. Based on the structure design parameters of snubber, distributed capacitance values of transmission line and power supply parameters, the overall configuration scheme of passive protection components in −1MV five-stage voltage system is analyzed. PubDate: 2023-09-25
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Abstract: Abstract Neutral Beam Injection (NBI), an auxiliary heating method in EAST, is evolving towards higher energy and longer pulses, leading to increased energy bombardment on the high-field side graphite protection tiles. This study explores the intricacies of theoretical injection energy and shine-through losses with respect to plasma density, NBI beam energy, and NBI pulse width. By evaluating the beam power distribution reaching the protective tiles, and subsequently assessing the temperature distribution on the tiles under varying pulse widths, a reliable comparison can be made between the simulated temperature results and experimental temperature data. To ensure the reliability of simulation results, the beam power distribution utilizes a two-dimensional Gaussian model and a particle model for contrast simulation. Similarly, the thermal deposition model employs a simplified uniform heating model, a analytic model and a finite element simulation model. As a reference, the experimentally measured data includes both infrared surface temperature measurements and buried thermocouple measurements. The objective of this study is to establish a correlation between key input variables and the protective tile temperature. By setting precise parameters, this research seeks to provide a predictive mechanism for thermal deposition, contributing to proactive overheating prevention and efficient adjustments to injection parameters and first-wall backend cooling metrics. PubDate: 2023-09-12
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Abstract: Abstract The operation of the Optimization of Liquid Metal Advanced Targets (OLMAT) facility began in April 2021 with the scientific objective of exposing liquid-metal plasma facing components (PFCs) to the particle and power fluxes provided by one of the hydrogen neutral beam injectors of the TJ-II stellarator. The system can deliver heat fluxes from 5 to 58 MW m−2 of high energy hydrogen neutral particles (≤ 33 keV) with fluxes up to 1022 m2 s−1 (containing an ion fraction ≤ 33% in some instances), pulsed operation of 30–150 ms duration and repetition rates up to 2 min−1. These characteristics enable OLMAT as a high heat flux (HHF) facility for PFC evaluation in terms of power exhaust capabilities, thermal fatigue and resilience to material damage. Additionally, the facility is equipped with a wide range of diagnostics that includes tools for analyzing the thermal response of the targets as well as for monitoring atomic/plasma physics phenomena. These include spectroscopy, pyrometry, electrical probing and visualization (fast and IR cameras) units. Such particularities make OLMAT a unique installation that can combine pure technological PFC research with the investigation of physical phenomena such as vapor shielding, thermal sputtering, the formation/characterization of plasma plumes with significant content of evaporated metal and the detection of impurities in front of the studied targets. Additionally, a myriad of surface characterization techniques as SEM/EDX for material characterization of the exposed PFC prototypes are available at CIEMAT. In this article, first we provide an overview of the current facility upgrade in which a high-power CW laser, that can be operated in continuous and pulsed modes (0.2–10 ms), dump and electrical (single Langmuir) probe embedded on the target surface have been installed. This laser operation will allow simulating more relevant heat loading scenarios such as nominal steady-state divertor heat fluxes (10–20 MW m−2 in continuous mode) and transients including ELM loading and disruption-like events (ms time scales and power densities up to GW m−2 range). The work later focuses on the more recent experimentation (2022 fall campaign) where a 3D printed Tungsten (W) Capillary Porous System (CPS) target, with approximated 30 μm pore size and a 37% porosity and filled with liquid tin. This porous surface was a mock-up of the PFC investigated in the ASDEX Upgrade divertor manipulator. The target composed with this element was eventually exposed to a sequence of shots with the maximum heat flux that OLMAT provides (58 ± 14 MWm−2). Key questions as resilience to dry-out and particle ejection of the liquid metal layer, its refilling, the induced damage/modification of the porous W matrix and the global performance of the component are addressed, attempting to shed light on the issues encountered with the PFC at tokamak scale testing. PubDate: 2023-09-04 DOI: 10.1007/s10894-023-00373-9
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Abstract: Abstract The liquid metal shield laboratory (LiMeS-Lab) will provide the infrastructure to develop, test, and compare liquid metal divertor designs for future fusion reactors. The main research topics of LiMeS-lab will be liquid metal interactions with the substrate material of the divertor, the continuous circulation and capillary refilling of the liquid metal during intense plasma heat loading and the retention of plasma particles in the liquid metal. To facilitate the research, four new devices are in development at the Dutch Institute for Fundamental Energy Research and the Eindhoven University of Technology: LiMeS-AM: a custom metal 3D printer based on powder bed fusion; LiMeS-Wetting, a plasma device to study the wetting of liquid metals on various substrates with different surface treatments; LiMeS-PSI, a linear plasma generator specifically adapted to operate continuous liquid metal loops. Special diagnostic protection will also be implemented to perform measurements in long duration shots without being affected by the liquid metal vapor; LiMeS-TDS, a thermal desorption spectroscopy system to characterize deuterium retention in a metal vapor environment. Each of these devices has specific challenges due to the presence and deposition of metal vapors that need to be addressed in order to function. In this paper, an overview of LiMeS-Lab will be given and the conceptual designs of the last three devices will be presented. PubDate: 2023-09-02 DOI: 10.1007/s10894-023-00379-3
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Abstract: Abstract The Lithium Evaporation EXperiment (LEEX) campaign on the Hybrid Illinois Device for Research and Applications (HIDRA) investigates helium retention effects induced by in-operando lithium evaporations into HIDRA. Lithium droplets were applied to tungsten samples and then exposed to a 600 s helium plasma at different distances, D, from the plasma edge. LEEX data has confirmed previous results at the University of Illinois Urbana-Champaign of in-operando lithium evaporations producing a low recycling regime for HIDRA helium plasmas and additionally proves the retained species is helium. The lithium evaporation from the D = 25 mm case had an 85.3% ± 1% increase in helium retention in the low recycling regime when compared to the steady state plasma of the LEEX control shot. Data presented substantiates previous helium retention claims and advances research surrounding liquid metal PFCs. A retention mechanism has not been identified, but further research utilizing HIDRA aims to investigate this. This study’s outcomes are thoroughly presented and provides additional justification for conducting further research on lithium’s behavior in fusion environments, given its substantial potential impact on the development of PFCs. PubDate: 2023-08-30 DOI: 10.1007/s10894-023-00383-7
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Abstract: Abstract The high magnetic field compact high-temperature superconducting tokamaks are characterized by high energy density, macroscopic instabilities, and high threshold values, which are beneficial to the operation of the device with high parameters, however, this also directly leads to a sharp increase in the heat loads of the divertors. Therefore, in this paper, with reference to the design parameters of SPARC, the TSC is used to simulate the double-null divertor sweeping configuration discharge process of a high magnetic field compact tokamak device (HFCT), and the heat flux distribution on the surface of the target at the time of divertor outer long-leg sweeping in the discharge flat-top stage is obtained, with the peak value of 23.2 MW/m2; both water-cooled monoblock and all-tungsten divertor configurations calculated using ANSYS, which concluded that the tungsten armor thickness d in the all-tungsten divertor configuration should not be greater than 2.5 mm to withstand a peak heat flux of 23.2 MW/m2; on the basis of this configuration, the temperature field of the divertor target is calculated for different sweeping fields and the feasibility of the all-tungsten divertor target model is analyzed. This study can be used as a reference for the design of divertor target in future HFCT. PubDate: 2023-08-28 DOI: 10.1007/s10894-023-00382-8
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Abstract: Abstract Self-healing liquid metal divertors (LMDs) based on the Capillary Porous Structure (CPS) concept are currently being considered among the possible solutions to the power exhaust problem in future fusion reactors. Indeed, the passive replenishment of the plasma-facing surface by capillary forces and the self-shielding of the target via vapor emission can potentially improve the divertor lifetime and its resilience to transient loads. On the other hand, the LMD target erosion can be significant due to evaporation and thermal sputtering, on top of physical sputtering, possibly leading to unacceptable core plasma dilution/power losses (for a low-Z/high-Z metal such as Li and Sn, respectively). For this reason, it is necessary to assess whether an LMD is compatible with an European DEMO (EU-DEMO) plasma scenario. This requires a self-consistent model of the impurity emission from the target, the plasma in both the scrape-off layer (SOL) and the core regions and the transport of impurities therein. In this paper, an an integrated modelling approach is proposed, which is based on SOLPS-ITER and includes its coupling with a target erosion model written in FreeFem++ and a core plasma model (ASTRA/STRAHL). An application of the coupled SOL-target model to simulate experiments performed in the Magnum-PSI linear plasma device with a CPS target filled with Li is also included to provide a first demonstration of the capabilities of the approach. Results are promising, being in good agreement (within a few degrees) with the measured target temperature distribution. In perspective, the modelling framework presented here will be applied to the EU-DEMO with an Sn divertor. PubDate: 2023-08-23 DOI: 10.1007/s10894-023-00377-5
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Abstract: Abstract To resolve technical issues associated with the plasma-facing components (PFCs) such as the divertor to be installed in a steady state magnetic fusion DEMO reactor, employing high-temperature metals such as tungsten for the surface component, the use of liquid metals (LMs) such as molten lithium has been proposed and evaluated as a possible resolution over the past two or so decades, using plasma confinement devices as well as laboratory-scale experimental facilities. The present work is intended to explore the effect of forced convection in liquid metals on the transport behavior of particles and heat from divertor plasma bombardment. Laboratory-scale experiments have been conducted, using GaInSn and molten lithium as the liquid metal targets to be exposed to steady-state plasmas and infrared irradiation. Data clearly indicate that electromagnetically induced convection can enhance particles and heat transport in these liquid metals, proof-of-principle data for convected LM-PFCs. PubDate: 2023-08-23 DOI: 10.1007/s10894-023-00375-7
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Abstract: Abstract The modelling of the edge lithium (Li) transport and heat flux deposition on divertor targets under different poloidal Li injections has been performed on EAST with the three-dimensional (3D) edge transport code EMC3-EIRENE. Four injected positions of Li impurity (including divertor and upstream injections) are investigated to check the Li density distribution and its impacts on the heat flux deposition profiles. The simulation results show that the Li injections near the strike points lead to small spatial amounts of Li ions compared to the upstream injections. The energy dissipation by Li impurity for different poloidal injections has been analysed by varying the upstream electron density. It is found that the increased upstream electron density leads to a slightly enhanced exhausted power of Li impurity for upstream injections while an evidently reduced energy loss for divertor injections. Moreover, the asymmetric distributions of heat flux deposition on target plates are obtained for divertor injections, while the symmetric distributions of heat flux are attained for upstream injections. PubDate: 2023-08-23 DOI: 10.1007/s10894-023-00376-6
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Abstract: Abstract According to the European Fusion Roadmap, to demonstrate the feasibility of fusion energy, it is necessary to build a comprehensive database of materials properties to be used in future fusion power plants. This is the objective of the IFMIF-DONES facility, which will drive a deuteron beam on a liquid lithium target to produce high energy neutron fluxes for irradiating candidate fusion materials. Among the several ongoing activities in the frame of the EUROfusion Early Neutron Source Work Package (WPENS) project, deterministic accident analyses play an important role, since they help identifying a set of reference accident scenarios and related safety class components. Some of these scenarios are being studied with the MELCOR-fusion code, an integrated engineering code which is able to perform thermal-hydraulic transient calculations. In this work, the MELCOR-fusion code has been applied to two potential accident scenarios involving the degradation of the primary lithium loop of IFMIF-DONES. A rupture in the Quench Tank and a break in the inlet nozzle to the Target Vacuum Chamber were postulated as the two initiating events followed by a lithium spill into the Test Cell room. The purpose of this study was to obtain different key metrics, such as the maximum pressure and temperature loads reached in the TC room, the amount of the leaked lithium mass, and the time for lithium solidification. The computed results will help identify the safety requirements to be applied to the final design of the TC room. PubDate: 2023-08-22 DOI: 10.1007/s10894-023-00378-4
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Abstract: Abstract We report sputtering yields of Li+, H−, O−, and OHx− ion species from an Li–O–H surface for H, D, He, Ne, and Ar ion irradiation at 45° incidence in the energy range of 30–2000 eV. A Li film was deposited on a stainless steel target using Li evaporators in the LTX-β vessel, using the LTX-β Sample Exposure Probe (SEP), which includes an ultrahigh vacuum suitcase for transferring targets without significant contamination from air exposure. The SEP was used to transfer the Li-coated target from LTX-β to a separate Sample Exposure Station (SES) to perform ion exposure measurements. The SEP was also used for characterization of the Li-coated target utilizing X-ray photoelectron spectroscopy in a different chamber, showing that the lithium film surface was oxidized. Ion exposures were performed using an electron cyclotron resonance plasma source in the SES. Sputtered/ejected species were sampled by a quadrupole mass spectrometer with capabilities for detecting positive and negative ions, and an energy filter for determining the mean kinetic energy of the ejected ion species. All ion irradiations caused Li+ ions to be ejected, while causing impurity ions such as H+, H−, O− and OH− to be ejected. Measured ion energies of Li+ ions from a Li–O–H surface suggested that the typical sheath potential on the divertor surface can trap sputtered Li+ ions, which were previously reported as ~ 60% of total sputtered Li species from Li targets (Allain and Ruzic in Nucl Fusion 42:202, 2002). Hence, our results for the sputtering yields of ejected ion species and their associated ion energies from a Li–O–H surface indicates that lithium sputtering is suppressed and impurity removal is enhanced due to the sheath potential at the divertor surface for fusion reactor applications. PubDate: 2023-08-18 DOI: 10.1007/s10894-023-00380-w
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Abstract: Abstract The influence of a low-pressure argon arc with a hot tungsten cathode on the thin tin film with a negative bias voltage applied during the plasma treatment was investigated to study the tin film removal from the sample surface. Samples were prepared on a stainless-steel substrate using DC magnetron sputtering and hybrid HiPIMS assisted with electron cyclotron wave resonance (ECWR). During treatment an optical emission spectroscopy was employed to detect and characterize the emission line of tin spectrum and the electron density and temperature were measured by Langmuir probe. Morphological study by a scanning electron microscope helped to gain insight to the mechanism of tin removal from the substrate. In addition, elemental compositions of tin layer before and after treatment was measured by an energy dispersive X-ray spectroscopy. We believe that this study contributes to finding a proper treatment for tin removal from plasma facing surfaces of tokamaks using tin in the liquid metal divertor. PubDate: 2023-08-18 DOI: 10.1007/s10894-023-00374-8
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