Study of electromagnetic disruption forces for plasma detachment measurements in DEMO (original) (raw)
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Influence of Plasma Composition on Divertor Detachment
2000
Abstract The phenomena that lead to the observed power and particle flux detachment have been studied with the code B2-Eirene for DIII-D Helium and Deuterium experiments. Contrary to the usual Deuterium experiments, a significant reduction of the power load to the divertor is observed in Helium discharges while the ion flux remains high. Modelling indicates that this is due to the longer ionisation mean free path of Helium, which can penetrate from the divertor into the bulk plasma with the consequent power loss.
Current understanding of divertor detachment: Experiments and modelling
2009
A qualitative as well as quantitative simulation of experimentally observed plasma parameters in the detached regime proves to be difficult for several tokamaks. A series of ohmic discharges have been performed in ASDEX Upgrade and DIII-D at as similar as possible plasma parameters and at different line averaged densities, n¯ e. The experimental data represent a set of well diagnosed discharges against which numerical simulations are compared. For the numerical modelling the fluid-code B2.
Plasma detachment in JET Mark I divertor experiments
Nuclear Fusion, 1998
The experimental characteristics of divertor detachment in the JET tokamak with the Mark I pumped divertor are presented for Ohmic, L-mode and ELMy H-mode experiments with the main emphasis on discharges with deuterium fuelling only. The range over which divertor detachment is observed for the various regimes as well as the influence of divertor configuration, direction of the toroidal field, divertor target material and active pumping on detachment will be described. The observed detachment characteristics such as the existence of a considerable electron pressure drop along the field lines in the scrape-off layer, and the compatibility of the decrease in plasma flux to the divertor plate with the observed increase of neutral pressure and the D α emission from the divertor region will be examined in the light of existing results from analytical and numerical models for plasma detachment. Finally, a method to evaluate the degree and window of detachment is proposed and all the observations of the JET Mark I divertor experiments summarised in the light of this new quantitative definition of divertor detachment.
Test of divertor materials under simulated ITER plasma disruption conditions at the GOL-3 facility
Journal of Nuclear Materials, 1994
The GOL-3 facility was used for exploratory plasma stream target experiments under conditions rather typical for the thermal quench phase of ITER tokamak plasma disruptions. The experiments allowed study of the properties of target plasmas formed from vaporized target materials in front of the target and determination of the target material erosion. Within 2 ps after the onset of the plasma stream a cloud of evaporated material is formed. The cloud expands along magnetic force lines with velocities around lo6 cm/s. Line radiation is observed from the target plasma corona, continuum radiation from the bulk of the cloud. The temperature in the plasma corona is about 1 eV. The black body temperature of the bulk cloud is below 0.5 eV. The erosion for graphite increases sharply upon reaching a threshold value of 1 MJ/m' for the energy density of the hot plasma stream and achieves very high values.
Measured plasma parameters in TdeV's closed poloidal divertors
Journal of Nuclear Materials, 1995
Measurements in TdeV in the double null configuration show strong asymmetries between the upper and lower divertors. The divertor to which the ion VB drift is directed has a higher density, which varies little with the central line-average density. The electron temperature also exhibits a poloidal variation which depends on the direction of the toroidal magnetic field. The divertor plasma shows a double peak structure which evolves as a function of the central density, with the outer peak varying more rapidly. We have measured flow reversal near the separatrix in the bottom divertor when the ion VB drift is directed away from the X-point. Divertor plate biasing allows us to control the flux of plasma in the SOL; under negative biasing there is a strong increase of the electron density in the active divertor, the one favoured by the E × B flow. Calculations of the energy deposition derived by flush-mounted probes agree well with measurements of the heat load derived from the temperature increase of the divertor tiles.
Insulated fixation system of plasma facing components to the divertor cassette in Eurofusion-DEMO
Fusion Engineering and Design, 2020
The design activities of an insulated Plasma Facing Components-Cassette Body (PFCs-CB) support has been carried out under the pre-conceptual design phase for Eurofusion-DEMO Work Package DIV-1 "Divertor Cassette Design and Integration"-Eurofusion Power Plant Physics & Technology (PPPT) program. The Eurofusion-DEMO divertor is a key in-vessel component with PFCs which directly interact with the plasma scrape-off layer. The PFCs have to cope with high heat loads, neutron irradiation and electromagnetic loads. The mechanical integrity of the PFCs and water cooling pipes can be jeopardized by high heat loads and by electromagnetic loads generated in a disruption event. In European-DEMO the possibility to estimate the heat load by measuring the relative thermocurrents is under investigation. In order to allow thermocurrents measurements, a divertor design option provides that PFCs are electrically insulated from CB. In this work authors aim to analyze the opportunity that the PFC-CB fixing system incorporates an electrical insulation system, thus acquiring also an important diagnostic role in the measurement of the thermocurrents and in the management of the current flows. The possible use of ceramic material (e.g. alumina) as the insulating layer between the support components is investigated.
He-cooled demo divertor: Design verification testing against mechanical impact loads
Fusion Engineering and Design, 2012
The design goal is to achieve a DEMO-relevant high heat flux of 10 MW/m 2. The reference design HEMJ (He-cooled modular divertor with multiple-jet cooling) uses small finger modules, which consist of a tungsten tile and a thimble made of tungsten alloy. Both components are connected by soldering. They are cooled by helium gas (10 MPa, 634 • C) impinging directly onto the inner heated surface of the thimble. One of the most difficult to predict incident events is the disruption that may damage the divertor structure by an extraordinary impact loading. This danger is particularly acute by the brittle property of the tungsten material, which is generally characterized by the DBTT. In this paper an estimate of the mechanical impact loading induced by electromagnetic forces during a disruption and an appropriate experimental setup are outlined, and the test results discussed.
IEEE Transactions on Magnetics, 1990
In the context of protection of next generation Tokamaks against plasma disruption events, numerical simulations of the electromagnetic/mechanical interactions on the segmented first wall (FW) system have been undertaken. Three-dimensional (3-D) computer codes have been applied for the resolution of the electromagnetic-type transient phenomena examining different finite element modelling techniques (shell and solid approaches) but considering decoupled levels of analysis (electrical and mechanical).
Shunt and Rogowski coil measurements on ASDEX Upgrade in support of DEMO detachment control
Fusion Engineering and Design, 2021
Detachment control in DEMO is a fundamental requirement to prevent damage to the plasma facing components. Thermo-currents flowing through the plasma facing components of the divertor cassette are driven by the thermoelectric voltage generated by the plasma temperature difference between the inner and outer target plates. Shunt and Rogowski coil measurements to measure thermo-currents for detachment control on ITER are planned. Thermo-current measurements on ASDEX Upgrade have been carried out in support of establishing designs considered suitable to measure thermocurrents in DEMO.
Vapor Shielding Effect on DEMO Divertor
Fusion Science and Technology, 2019
The impact of the edge-localized modes (ELMs) on the tungsten divertor erosion by taking into account the screening effect of vapor shielding is analyzed for DEMO steady-state operation condition. The evaluation of tungsten ablation, energy radiation, and absorption by divertor plate due to a single ELM impact is calculated by using a model of vapor shielding inserted in the MEMOS code. The effect of repetitive ELM impact and the tungsten melt layer formation is described by using the model of W monoblock with a compliance layer of Cu alloy between the W and EUROFER water cooling tube. It is shown that the vapor plasma shielding results in saturation of the single ELM energy accumulated by the divertor plate and that the saturation level depends on the ELM duration. The ablation thickness can reach about 0.01 µm. The total number of ablated particles is rather critical for the shielding formation, and the lifetime of the divertor plate depends strongly on this effect.