Search results for "TOKAMAK"
showing 10 items of 54 documents
Magnetic configuration effects on the Wendelstein 7-X stellarator
2018
The two leading concepts for confining high-temperature fusion plasmas are the tokamak and the stellarator. Tokamaks are rotationally symmetric and use a large plasma current to achieve confinement, whereas stellarators are non-axisymmetric and employ three-dimensionally shaped magnetic field coils to twist the field and confine the plasma. As a result, the magnetic field of a stellarator needs to be carefully designed to minimize the collisional transport arising from poorly confined particle orbits, which would otherwise cause excessive power losses at high plasma temperatures. In addition, this type of transport leads to the appearance of a net toroidal plasma current, the so-called boot…
Analysis of equations arising in gyrotron theory
2012
The gyrotron is a microwave source whose operation is based on the stimulated cyclotron radiation of electrons oscillating in a static magnetic field. Powerful gyrotrons can be used to heat nuclear fusion plasma. In addition, they have found a wide utility in plasma diagnostics, plasma chemistry, radars, extra-high-resolution spectroscopy, high-temperature processing of materials, medicine, etc. However, the main application of gyrotrons is in electron cyclotron resonance heating in tokamaks and stellarators. Equations describing gyrotron operation are ordinary differential equations and Schrödinger type partial differential equations. The present paper provides a survey of the analytical a…
A computational procedure for the investigation of whipping effect on ITER High Energy Piping and its application to the ITER divertor primary heat t…
2015
Abstract The Tokamak Cooling Water System of nuclear facility has the function to remove heat from plasma facing components maintaining coolant temperatures, pressures and flow rates as required and, depending on thermal-hydraulic requirements, its systems are defined as High Energy Piping (HEP) because they contain fluids, such as water or steam, at a pressure greater than or equal to 2.0 MPa and/or at a temperature greater than or equal to 100 °C, or even gas at pressure above the atmospheric one. The French standards contemplate the need to consider the whipping effect on HEP design. This effect happens when, after a double ended guillotine break, the reaction force could create a displa…
Steady state and transient thermal–hydraulic analyses on ITER divertor module
2005
Abstract One of the most challenging components of ITER is the divertor devoted at controlling the characteristics of the plasma boundary, exhausting the α particles and reducing the impurities in the plasma. The thermal–hydraulic design of the divertor is particularly, demanding because of the high heat loads and the cooling flow margin in the plasma-facing components (PFCs). The pressure drop is limited by the pumping power and also avoiding the risk of reaching critical heat flux (CHF). Furthermore, for maintenance operation foreseen, each single divertor cassette should be drained and dried before withdrawing it out from the vacuum vessel. To address these requirements, European Fusion …
Analysis of metallic impurity content by means of VUV and SXR diagnostics in hybrid discharges with hot-spots on the JET-ITER-like wall poloidal limi…
2019
In preparation for the upcoming JET D-T campaign, great effort has been devoted during the 2015-2016 JET campaigns with the ITER-like wall (ILW) to the extension of the high performance H-mode phase in baseline and hybrid scenarios. Hybrid discharges were the only ones that have been stopped by the real-time vessel protection system due hot-spot formation on the outboard poloidal limiter. Generation of hot-spots was linked to the application of high neutral beams injection and ion cyclotron resonance heating (ICRH) power. In tokamaks with high-Z plasma components, the use of ICRH heating is also accompanied by an increased metallic impurity content. Simultaneous control of hot-spot temperat…
Comparison of tritium measurement techniques for a laser cleaned JET tile
2014
Abstract Over the last 7–8 years, two quantitative analyzing methods—accelerator mass spectrometry (AMS) and full combustion (FC) followed by scintillation detection have been applied for determining the tritium activity concentrations in JET divertor tiles. These methods have two main differences – the range of detection and the spatial resolution – and are thus complementary. However, these differences can also complicate the comparison of the two techniques for typical JET divertor samples. Therefore a cross comparison exercise for tritium measurements was performed between the two methods using specially produced identical standard samples. The cross comparison measurements were perform…
Transient DC-Arc Voltage Model in the Hybrid Switch of the DTT Fast Discharge Unit
2021
The focus of this work is the transient modelling of the DC-arc voltage on a Hybrid Switch (a mechanical switch in parallel with a static switch) of a key protection component called Fast Discharge Unit (FDU) in the Divertor Tokamak Test (DTT). The DTT facility is an experimental tokamak in advanced design and realization phase, which will be built in the ENEA Research Centre in Frascati (Italy). The FDU allows the safe discharge of the Toroidal Field (TF) superconducting magnets when a quench is detected or a failure occurs in the power supply or in the cryogenic system. In this work, the arc conductance of the mechanical By-Pass Switch (BPS) of the Hybrid Switch is modelled using the well…
Structural assessment of a whole toroidal sector of the HELIAS 5-B breeding blanket
2021
Abstract The European roadmap for the realization of fusion energy considers the stellarator line as a possible long-term alternative to a tokamak DEMO. In this context, from the plasma physics standpoint, the most promising option is a five-field period power plant called HELIcal-axis Advanced Stellarator (HELIAS) 5-B. In order to allow the electricity production, the HELIAS 5-B reactor must be endowed with a breeding blanket (BB). Hence, in this paper, the advancements in the HELIAS 5-B BB design are reported. In particular, the structural assessment of a whole BB period, extending along toroidal direction for 72 °, is depicted. A geometric configuration encompassing dummy BB segments has…
Preliminary structural assessment of the HELIAS 5-B breeding blanket
2019
Abstract The European Roadmap to the realisation of fusion energy, carried out by the EUROfusion consortium, considers the stellarator concept as a possible long-term alternative to a tokamak fusion power plant. To this purpose a pivotal issue is the design of a HELIcal-axis Advanced Stellarator (HELIAS) machine equipped with a tritium Breeding Blanket (BB), considering the achievements and the design experience acquired in the pre-conceptual design phase of the tokamak DEMO BB. Therefore, within the framework of EUROfusion Work Package S2 R&D activity, a research campaign has been launched at KIT. The scope of the research has been the determination of a preliminary BB segmentation scheme …
Overview of the JET results in support to ITER
2017
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 m…