| 1. |
- Rubel, Marek J., et al.
(författare)
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Molybdenum limiters for Extrap-T2 upgrade : surface properties and high heat flux testing
- 2000
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Ingår i: Fusion engineering and design. - 0920-3796 .- 1873-7196. ; 49, s. 323-329
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Tidskriftsartikel (refereegranskat)abstract
- A vacuum vessel, a conductive copper shell and plasma-facing components of the Extrap-T2 device, a medium size reversed field pinch, are under major rebuild. The new machine is equipped with an array of 180 molybdenum limiters, which will be exposed to power loads of about 30 MW m(-2), but higher loads can not be excluded. Prior to the limiters' installation in the vessel, they were tested under high heat loads in the JUDITH electron beam facility in order to assess the possible damage to the surface and to the bulk of the material. A test limiter was irradiated with a beam of increasing power density from 15 to 1500 MW m(-2). Surface characterization was performed before and after the irradiation using electron and optical microscopy, energy dispersive X-ray spectroscopy, enhanced proton scattering and laser profilometry. Metallography studies were performed for the irradiated areas. The irradiation induced the change in surface morphology, e.g. surface melting and re-crystallization of grains, only following the 10 ms long pulses with the absorbed power density approaching 1500 MW m(-2).
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| 2. |
- Tanabe, T., et al.
(författare)
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Material mixing on W/C twin limiter in TEXTOR-94
- 2000
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Ingår i: Fusion engineering and design. - 0920-3796 .- 1873-7196. ; 49, s. 355-362
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Tidskriftsartikel (refereegranskat)abstract
- In order to investigate the effect of mutual contamination between tungsten (W) and carbon
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| 3. |
- Agredano Torres, Manuel, et al.
(författare)
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Coils and power supplies design for the SMART tokamak
- 2021
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 168, s. 112683-112683
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Tidskriftsartikel (refereegranskat)abstract
- A new spherical tokamak, the SMall Aspect Ratio Tokamak (SMART), is currently being designed at the University of Seville. The goal of the machine is to achieve a toroidal field of 1 T, a plasma current of 500 kA and a pulse length of 500 ms for a plasma with a major radius of 0.4 m and minor radius of 0.25 m. This contribution presents the design of the coils and power supplies of the machine. The design foresees a central solenoid, 12 toroidal field coils and 8 poloidal field coils. Taking the current waveforms for these set of coils as starting point, each of them has been designed to withstand the Joule heating during the tokamak operation time. An analytical thermal model is employed to obtain the cross sections of each coil and, finally, their dimensions and parameters. The design of flexible and modular power supplies, based on IGBTs and supercapacitors, is presented. The topologies and control strategy of the power supplies are explained, together with a model in MATLAB Simulink to simulate the power supplies performance, proving their feasibility before the construction of the system.
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| 4. |
- Baron-Wiechec, A., et al.
(författare)
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Thermal desorption spectrometry of beryllium plasma facing tiles exposed in the JET tokamak
- 2018
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Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 133, s. 135-141
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Tidskriftsartikel (refereegranskat)abstract
- The phenomena of retention and de-trapping of deuterium (D) and tritium (T) in plasma facing components (PFC) and supporting structures must be understood in order to limit or control total T inventory in larger future fusion devices such as ITER, DEMO and commercial machines. The goal of this paper is to present details of the thermal desorption spectrometry (TDS) system applied in total fuel retention assessment of PFC at the Joint European Torus (JET). Examples of TDS results from beryllium (Be) wall tile samples exposed to JET plasma in PFC configuration mirroring the planned ITER PFC is shown for the first time. The method for quantifying D by comparison of results from a sample of known D content was confirmed acceptable. The D inventory calculations obtained from Ion Beam Analysis (IBA) and TDS agree well within an error associated with the extrapolation from very few data points to a large surface area.
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| 5. |
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| 6. |
- Batistoni, P., et al.
(författare)
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Technical preparations for the in-vessel 14 MeV neutron calibration at JET
- 2017
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Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 117, s. 107-114
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Tidskriftsartikel (refereegranskat)abstract
- The power output of fusion devices is measured from their neutron yields which relate directly to the fusion yield. In this paper we describe the devices and methods that have been prepared to perform a new in situ 14 MeV neutron calibration at JET in view of the new DT campaign planned at JET in the next years. The target accuracy of this calibration is 10% as required for ITER, where a precise neutron yield measurement is important, e.g., for tritium accountancy. In this paper, the constraints and early decisions which defined the main calibration approach are discussed, e.g., the choice of 14 MeV neutron source and the deployment method. The physics preparations, source issues, safety and engineering aspects required to calibrate directly the JET neutron detectors are also discussed. The existing JET remote-handling system will be used to deploy the neutron source inside the JET vessel. For this purpose, compatible tooling and systems necessary to ensure safe and efficient deployment have been developed. The scientific programme of the preparatory phase is devoted to fully characterizing the selected 14 MeV neutron generator to be used as the calibrating source, obtain a better understanding of the limitations of the calibration, optimise the measurements and other provisions, and to provide corrections for perturbing factors (e.g., anisotropy of the neutron generator, neutron energy spectrum dependence on emission angle). Much of this work has been based on an extensive programme of Monte-Carlo calculations which provide support and guidance in developing the calibration strategy.
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| 7. |
- Batistoni, P., et al.
(författare)
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Technological exploitation of Deuterium-Tritium operations at JET in support of ITER design, operation and safety
- 2016
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Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 109, s. 278-285
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Tidskriftsartikel (refereegranskat)abstract
- Within the framework of the EUROfusion programme, a work-package of technology projects (WPJET3) is being carried out in conjunction with the planned Deuterium-Tritium experiment on JET (DTE2) with the objective of maximising the scientific and technological return of DT operations at JET in support of ITER. This paper presents the progress since the start of the project in 2014 in the preparatory experiments, analyses and studies in the areas of neutronics, neutron induced activation and damage in ITER materials, nuclear safety, tritium retention, permeation and outgassing, and waste production in preparation of DTE2.
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| 8. |
- Batistoni, P., et al.
(författare)
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The JET technology program in support of ITER
- 2014
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 89:7-8, s. 896-900
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Tidskriftsartikel (refereegranskat)abstract
- This paper presents an overview of the current and planned technological activities at JET in support of ITER operation and safety. The scope is very broad and it ranges from analysis of components from the ITER-like Wall (ILW) to determine material erosion and deposition, dust generation and fuel retention to neutronics measurements and analyses. Preliminary results are given of the post-mortem analyses of samples exposed to JET plasmas during the first JET-ILW operation in 2011-2012, and retrieved during the following in-vessel intervention. JET is the only fusion machine capable of producing significant neutron yields, up to nearly 10(19) n/s (14.1 MeV) in DT operations. Recently, the technological potential of a new DT campaign at JET in support of ITER has been explored and the outcome of this assessment is presented. The expected 14 MeV neutron yield, the use of tritium, the preparation and implementation of safety measures will provide a unique occasion to gain experience in several ITER relevant technological areas. A number of projects and experiments to be conducted in conjunction with the DT operation have been identified and they are described in this paper.
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| 9. |
- Biel, W., et al.
(författare)
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Development of a concept and basis for the DEMO diagnostic and control system
- 2022
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Ingår i: Fusion engineering and design. - : Elsevier. - 0920-3796 .- 1873-7196. ; 179
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Tidskriftsartikel (refereegranskat)abstract
- An initial concept for the plasma diagnostic and control (D&C) system has been developed as part of European studies towards the development of a demonstration tokamak fusion reactor (DEMO). The main objective is to develop a feasible, integrated concept design of the DEMO D&C system that can provide reliable plasma control and high performance (electricity output) over extended periods of operation. While the fusion power is maximized when operating near to the operational limits of the tokamak, the reliability of operation typically improves when choosing parameters significantly distant from these limits. In addition to these conflicting requirements, the D&C development has to cope with strong adverse effects acting on all in vessel components on DEMO (harsh neutron environment, particle fluxes, temperatures, electromagnetic forces, etc.). Moreover, space allocation and plasma access are constrained by the needs for first wall integrity and optimization of tritium breeding. Taking into account these boundary conditions, the main DEMO plasma control issues have been formulated, and a list of diagnostic systems and channels needed for plasma control has been developed, which were selected for their robustness and the required coverage of control issues. For a validation and refinement of this concept, simulation tools are being refined and applied for equilibrium, kinetic and mode control studies.
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| 10. |
- Biel, W., et al.
(författare)
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Diagnostics for plasma control - : From ITER to DEMO
- 2019
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Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 146:A, s. 465-472
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Tidskriftsartikel (refereegranskat)abstract
- The plasma diagnostic and control (D&C) system for a future tokamak demonstration fusion reactor (DEMO) will have to provide reliable operation near technical and physics limits, while its front-end components will be subject to strong adverse effects within the nuclear and high temperature plasma environment. The ongoing developments for the ITER D&C system represent an important starting point for progressing towards DEMO. Requirements for detailed exploration of physics are however pushing the ITER diagnostic design towards using sophisticated methods and aiming for large spatial coverage and high signal intensities, so that many front-end components have to be mounted in forward positions. In many cases this results in a rapid aging of diagnostic components, so that additional measures like protection shutters, plasma based mirror cleaning or modular approaches for frequent maintenance and exchange are being developed. Under the even stronger fluences of plasma particles, neutron/gamma and radiation loads on DEMO, durable and reliable signals for plasma control can only be obtained by selecting diagnostic methods with regard to their robustness, and retracting vulnerable front-end components into protected locations. Based on this approach, an initial DEMO D&C concept is presented, which covers all major control issues by signals to be derived from at least two different diagnostic methods (risk mitigation).
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