Search results for "fusion power"
showing 10 items of 78 documents
Advancements in the Helium-Cooled Pebble Bed Breeding Blanket for the EU DEMO: Holistic Design Approach and Lessons Learned
2019
The helium-cooled pebble bed (HCPB) blanket is one of the two concepts proposed as a driver blanket for the European Union Demonstration Fusion Power Reactor (EU DEMO). In contrast to past conceptual design studies, in the frame of the current Power Plant Physics and Technology of the EUROfusion Consortium, the ongoing EU DEMO preconceptual design activities have adopted a holistic and integrated (i.e., systems engineering) design approach. As a consequence of this new approach, many interfaces and requirements have been identified, some of them driving the design of the blankets. This paper shows the advancements in the HCPB breeding blanket and describes the lessons learned after implemen…
Magnetic field effects on tritium release from neutron-irradiated beryllium pebbles
2007
The effects of temperature, magnetic field (MF), and ionizing radiation on the release of tritium from the Be pebbles irradiated in the BERYLLIUM experiment in 1994 in Petten, The Netherlands (irradiation neutron fluence 1.24×10 25 m -2 , irradiation temperature 780 K, and 3 H content 7 appm) were investigated in this study. Simultaneous action of these factors corresponds to the real operating conditions of the blanket of a fusion reactor. The total amount of tritium in a separate pebble, the chemical forms of localized tritium (T 0 , T 2 , and T + ), and the tritium distribution in the pebble volume were determined by a lyomethod (dissolution). Thermoannealing experiments were performed a…
Tritium localisation and release from the ceramic pebbles of breeder
2004
Magnetic field (MF) effects on the radiolysis and tritium release from Li4SiO4 (FZK) and Li2TiO3 (CEA) ceramic pebbles were investigated. The tritium chemical forms in Li4SiO4 were estimated by means of lyomethods. In the case of the neutron fluence Fn 6 10 18 nm � 2 , the tritium is mostly in the T þ form, but in the case of Fn � 10 25 nm � 2 , the T þ form accounts for 86–95% of the tritium. A high subsurface concentration of tritium is characteristic of a separate pebble and correlates with the distribution of radiation-induced defects. The MF increases the radiolysis of Li4SiO4 by 20–25%. Irradiation with electrons to 1000 MGy at 1200 K increases the grain size by 5–10%, decreasing the …
Preliminary Thermo-Mechanical Design of the Once Through Steam Generator and Molten Salt Intermediate Heat Exchanger for EU DEMO
2020
The European DEMOnstration power plant (DEMO) is considered to be the nearest-term fusion reactor capable of producing several hundred MWs of net electricity, operating with a closed tritium fuel-cycle (achieving the tritium self-sufficiency), and qualifing technological solutions for a fusion power plant with different breeding blanket (BB) concepts that are under investigation. The BB is a key component for the development of the DEMO plant design and, in particular, of those systems having the responsibility to remove the plasma-generated thermal power and its conversion in the electrical energy. This study deals with the preliminary thermo-mechanical design of the heat exchangers and st…
Liquid gallium jet–plasma interaction studies in ISTTOK tokamak
2009
Abstract Liquid metals have been pointed out as a suitable solution to solve problems related to the use of solid walls submitted to high power loads allowing, simultaneously, an efficient heat exhaustion process from fusion devices. The most promising candidate materials are lithium and gallium. However, lithium has a short liquid state temperature range when compared with gallium. To explore further this property, ISTTOK tokamak is being used to test the interaction of a free flying liquid gallium jet with the plasma. ISTTOK has been successfully operated with this jet without noticeable discharge degradation and no severe effect on the main plasma parameters or a significant plasma conta…
On the use of tin–lithium alloys as breeder material for blankets of fusion power plants
2000
Abstract Tin–lithium alloys have several attractive thermo-physical properties, in particular high thermal conductivity and heat capacity, that make them potentially interesting candidates for use in liquid metal blankets. This paper presents an evaluation of the advantages and drawbacks caused by the substitution of the currently employed alloy lead–lithium (Pb–17Li) by a suitable tin–lithium alloy: (i) for the European water-cooled Pb–17Li (WCLL) blanket concept with reduced activation ferritic–martensitic steel as the structural material; (ii) for the European self-cooled TAURO blanket with SiCf/SiC as the structural material. It was found that in none of these blankets Sn–Li alloys woul…
Behaviour of neutron irradiated beryllium during temperature excursions up to and beyond its melting temperature
2015
Abstract Beryllium pebble behaviour has been studied regarding the accidental operation conditions of tritium breeding blanket of fusion reactors. Structure evolution, oxidation and thermal properties have been compared for nonirradiated and neutron irradiated beryllium pebbles during thermal treatment in a temperature range from ambient temperature to 1600 K. For neutron irradiated pebbles tritium release process was studied. Methods of temperature programmed tritium desorption (TPD) in combination with thermogravimetry (TG) and temperature differential analysis (TDA), scanning electron microscopy (SEM) in combination with Energy Dispersive X-ray analysis (EDX) have been used. It was found…
Technical challenges in the construction of the steady-state stellarator Wendelstein 7-X
2013
The next step in the Wendelstein stellarator line is the large superconducting device Wendelstein 7-X, currently under construction in Greifswald, Germany. Steady-state operation is an intrinsic feature of stellarators, and one key element of the Wendelstein 7-X mission is to demonstrate steady-state operation under plasma conditions relevant for a fusion power plant. Steady-state operation of a fusion device, on the one hand, requires the implementation of special technologies, giving rise to technical challenges during the design, fabrication and assembly of such a device. On the other hand, also the physics development of steady-state operation at high plasma performance poses a challeng…
Ab initio modelling of the initial stages of the ODS particle formation process
2018
Abstract Oxide-Dispersion Strengthened (ODS) steels with Y2O3 nanoparticles are promising structural materials for fision and future fusion reactors. A large number of experimental as well as theoretical studies provided valuable information on the ODS particle formation process. However, some important details of this process still remain unexplained. We present the results of ab initio VASP calculations of the initial steps of the ODS particle formation. At these steps Y solute atoms are stabilized in the Fe lattice by vacancies, which create a basis for the future growth of Y2O3-particle. Interaction of multiple vacancies and solution Y and O atoms has been studied in various combination…
On the effects of the supporting frame on the radiation-induced damage of HCLL-TBM structural material
2007
Within the European Fusion Technology Programme, research activities have been conducted on the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept with the aim of manufacturing a Test Blanket Module (TBM) to be irradiated in ITER. HCLL-TBM is planned to be located in an ITER equatorial port, housed inside an AISI 316 stainless steel-supporting frame. Since that frame has been designed to provide two positions separated by a dividing plate and the HCLL-TBM is expected to fill one of them, its nuclear response could vary depending on the filling status of the other position and on the plate thickness. A parametric study has been carried out to investigate the potential effects on the …