The paper describes the progress of the studies on first wall and liquid breeder blankets for tri... more The paper describes the progress of the studies on first wall and liquid breeder blankets for tritium production in the Next European Torus (NET). Two concepts of first wall/blanket segments are described, using 17Li83Pb as breeder and water as coolant. In both concepts the first wall is integrated in a steel box enveloping the breeder units which are cylindrical vessels
2007 IEEE 22nd Symposium on Fusion Engineering, 2007
A portable reaming tool has been developed, manufactured and successfully tested to accurately re... more A portable reaming tool has been developed, manufactured and successfully tested to accurately ream holes in flanged connections for the superconducting coil support structures of Wendelstein 7-X stellarator in areas where clean conditions are required to meet high vacuum requirements. The tool operates in spaces with limited access and is able to ream without lubricant. It is composed of a
The driving concepts of systems integration, based on assembly, disassembly and maintenance requi... more The driving concepts of systems integration, based on assembly, disassembly and maintenance requirements which define the mechanical configuration of INTOR (a world-wide conceptual study of an experimental Tokamak-type power fusion reactor of the next generation), are presented as the starting point for the studies carried out in this field at JRC-Ispra. Complementary new developments recently incorporated into the European version of INTOR, referred to as INTOR-NET, are described in detail and compared with the original concepts. The aim in INTOR-NET has been to reduce the physical size of the reactor while retaining similar plasma parameters. New systems integration and mechanical configuration concepts are introduced which can be used in future investigations for the NET design as alternative options. Further reductions in reactor and/or improvements in the maintenance approach appear possible.
The Helium Cooled Lithium Lead (HCLL) and the Helium Cooled Pebble Bed (HCPB) Blanket are the ref... more The Helium Cooled Lithium Lead (HCLL) and the Helium Cooled Pebble Bed (HCPB) Blanket are the reference concepts in the European Breeding Blanket Programme for the DEMO design and for the related long term R&D. Recently, a similar design for both concepts has been developed, in particular both concepts use helium coolant and RAFM steel EUROFER as structural material. In
In the European Union, the basic manufacturing technologies have been selected for the two refere... more In the European Union, the basic manufacturing technologies have been selected for the two reference DEMO breeding blanket concepts: the helium cooled pebble bed and the helium cooled lithium lead. These technologies have been tested in the past years on small-scale samples and mock-ups, are presently under further development, and are now being adapted to the evolution of the two
The in-vessel components of the WENDELSTEIN 7-X stellarator consist of the divertor components an... more The in-vessel components of the WENDELSTEIN 7-X stellarator consist of the divertor components and the wall protection with its internal cooling supply. The main components of the open divertor are the vertical and horizontal target plates which form the pumping gap, the cryo-vacuum pumps and the control coils. The divertor volume is closed by graphite shielded baffle modules and with divertor closures. All these components are designed to be actively water-cooled. For the first commissioning phase planned in 2014, an inertial-cooled test divertor will be installed instead of the actively water-cooled high heat flux divertor. The wall protection consists of graphite-protected heat shields in the higher loaded areas and stainless steel panels in the lower loaded regions. The wall protection cooling circuits are connected through 80 supply-ports via so-called "plug-ins". It is envisaged to protect the diagnostic ports by paneltype port-liners. Special graphite-shielded port liners are used on the diagnostic injector and the neutral beam injector ports. The in-vessel components are mainly manufactured and tested at the Max-Planck-Institute für Plasmaphysik in its Garching workshop. Panels, high heat flux target elements and control coils are delivered by industrial partners. Manufacturing of the KiP ("Komponenten im Plasmagefäß") is in plan. Delivery of the components will be in time.
This article in an outline of the R&D progress on low Z materials and plasma facing components fo... more This article in an outline of the R&D progress on low Z materials and plasma facing components for NET. It covers the following topics: (1) neutron irradiation effects in low Z materials; (2) the evaluation of effects on plasma facing materials from off-normal operation events, i.e. disruption heat loads and runaway-electron events; (3) manufacture of first wall mock-ups with and without carbon protection tiles and simulation of the cyclic heat loads during normal operation; (4) development of design options for the divertor and a stepwise approach from materials screening tests to medium size mock-up tests for the design and development of the divertor.
... for 100 preparatory cycles and 900 cycles of 30 s heating and 30 s cooling time. At the end o... more ... for 100 preparatory cycles and 900 cycles of 30 s heating and 30 s cooling time. At the end of the 1000 cycles, the surface and the Be/DS-Cu joint of the mock up did not show any damage due to the fatigue test. Article Outline. ...
ABSTRACT In order to improve the tritium release capability of beryllium in helium cooled ceramic... more ABSTRACT In order to improve the tritium release capability of beryllium in helium cooled ceramic breeder (HCPB) type blankets, beryllium tablets with small grain sizes ranging from 0.4 to 20 μm and 15 mm diameter and a few millimetres height have been produced by a wet ball-milling process with subsequent hot pressing. Depending on process parameters, about 0.4–1.5 kg Be powder can be produced per day. X-ray investigations, SEM and EDX confirmed among others that the impurity concentrations could be kept very close to the raw material. As grain coarsening of Be is at a given temperature besides pressure and possible grain surface layers a sensitive function of grain size, Be tablets have been subjected to isochronal annealing for 2 h between 500 and 1000 °C in high vacuum. While annealing up to 800 °C did not change the average grain size of about 5 μm on the basis of the initial datasets, grain coarsening occurs clearly during 900 °C annealing.
One of the key issues of the European Helium Cooled Pebble Bed blanket is the behaviour under irr... more One of the key issues of the European Helium Cooled Pebble Bed blanket is the behaviour under irradiation of beryllium pebbles, which have the function of neutron multiplier. An intense production of helium occurs in-pile, as well as a non negligible generation of tritium. Helium bubbles induce swelling and a high tritium inventory is a safety issue. Extensive studies for a better understanding, characterisation and modelling of the behaviour of helium and tritium in irradiated beryllium pebbles are being carried out, with the final aim to enable a reliable prediction of gas release and swelling in the full range of operating and accidental conditions of a Fusion Power Reactor. The general strategy consists in integrating studies on macroscopic phenomena (gas release) with the characterisation of corresponding microscopic diffusion phenomena (bubble kinetics) and the assessment of some fundamental diffusion parameter for the models (gas atomic diffusion coefficients). The present work gives a summary of the latest achievements in this context. By an inverse analysis of experimental out-of-pile gas release from weakly irradiated pebbles, coupled to the study of the characteristics of bubble population, it has been possible to assess the thermal diffusion coefficients of helium and tritium in and to improve and validate the classical model of gas precipitation into bubbles inside the grain. The improvement of the description of gas atomic diffusion and precipitation is the first step to enable a more reliable prediction of gas release.
Equipos Nucleares S.A. as one of the contractors for the Wendelstein 7-X Stellarator project is c... more Equipos Nucleares S.A. as one of the contractors for the Wendelstein 7-X Stellarator project is currently manufacturing the coil support structure (CSS), which is a stainless steel ring-shaped structure formed by five equal sectors (modules), each of them made up from two symmetric half-modules, everything joined together by way of bolts and super-bolts. This CSS works as the central support for both planar and non-planar coils. The complicate shape of the CSS allows the ports to pass through the structure and reach their place into the vacuum vessel.
The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Pla... more The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Plasma Facing Components (PFC) which focused on the following main tasks: (a) the definition of basic design concepts for the First Wall (FW) and Divertor Plates (DP), (b) the analysis of the performance and likely lifetime of these PFC designs including the identification of major critical issues, (c) the start of R&D work giving already first results, and the definition of the required further R&D program to support the contemplated ITER Engineering Design Activity (EDA).
The long term European materials R&D prog... more The long term European materials R&D programme is mainly focused on the development of the following DEMO relevant materials: structural (RAFM steel called EUROFER), plasma facing (W alloy), breeder and neutron multiplier (Pb–17Li, Li4SiO4, Li2TiO3 and Be). Materials like ODS and SiCf/SiC composite are also investigated for advanced fusion reactor concepts. In addition, the materials programme includes activities in nuclear
The operation of W7-X stellarator for pulse length up to 30min with 10MW input power requires a f... more The operation of W7-X stellarator for pulse length up to 30min with 10MW input power requires a full set of actively water-cooled plasma facing components. From the lower thermally loaded area of the wall protection system designed for an averaged load of 100kW/m2 to the higher loaded area of the divertor up to 10MW/m2, various design and technological solutions have
The electrodes for the Wendelstein 7-X glow discharge system have been designed, tested and manuf... more The electrodes for the Wendelstein 7-X glow discharge system have been designed, tested and manufactured. The compact design relies on a cooled housing, integrated into the first wall cooling system, and a calotte-shaped graphite anode. The new mounting concept avoids the need of active cooling of the anode due to an improved thermal conduction. Comprehensive tests of a prototype electrode
All in vessel components (IVCs) of W7-X are actively cooled. Inside the plasma vessel about 4km o... more All in vessel components (IVCs) of W7-X are actively cooled. Inside the plasma vessel about 4km of pipes will be installed, supplying water to the IVC. 226 cooling circuits with 78 variants are necessary. The cooling circuits enter the cryostat and the plasma vessel through ad hoc flanged penetrations called “plug-ins”, which provide for the vacuum boundary between the plasma
Significant improvements of the ITER blanket were possible with the better definition of the ther... more Significant improvements of the ITER blanket were possible with the better definition of the thermal and electromagnetic effects. A supporting double wall backplate with reinforcement about the ports has been defined. The module has been simplified and the intrusion of the connections reduced. The attachment has been assessed by system analyses. The cooling system is configured to simplify the leak detection. 0-7803-4226-7/98/$10.00 0 1998 IEEE
The ITER General Design Requirements call for a design allowing the in-situ repair of protective ... more The ITER General Design Requirements call for a design allowing the in-situ repair of protective armours and, if possible, the separate replacement of FW sub-components, in order to minimise intervention time and radioactive waste generation. The technologies necessary to implement this maintenance strategy and the relevant experimental results of the on going ITER R&D program specifically tailored for their development are discussed in this paper .
The paper describes the progress of the studies on first wall and liquid breeder blankets for tri... more The paper describes the progress of the studies on first wall and liquid breeder blankets for tritium production in the Next European Torus (NET). Two concepts of first wall/blanket segments are described, using 17Li83Pb as breeder and water as coolant. In both concepts the first wall is integrated in a steel box enveloping the breeder units which are cylindrical vessels
2007 IEEE 22nd Symposium on Fusion Engineering, 2007
A portable reaming tool has been developed, manufactured and successfully tested to accurately re... more A portable reaming tool has been developed, manufactured and successfully tested to accurately ream holes in flanged connections for the superconducting coil support structures of Wendelstein 7-X stellarator in areas where clean conditions are required to meet high vacuum requirements. The tool operates in spaces with limited access and is able to ream without lubricant. It is composed of a
The driving concepts of systems integration, based on assembly, disassembly and maintenance requi... more The driving concepts of systems integration, based on assembly, disassembly and maintenance requirements which define the mechanical configuration of INTOR (a world-wide conceptual study of an experimental Tokamak-type power fusion reactor of the next generation), are presented as the starting point for the studies carried out in this field at JRC-Ispra. Complementary new developments recently incorporated into the European version of INTOR, referred to as INTOR-NET, are described in detail and compared with the original concepts. The aim in INTOR-NET has been to reduce the physical size of the reactor while retaining similar plasma parameters. New systems integration and mechanical configuration concepts are introduced which can be used in future investigations for the NET design as alternative options. Further reductions in reactor and/or improvements in the maintenance approach appear possible.
The Helium Cooled Lithium Lead (HCLL) and the Helium Cooled Pebble Bed (HCPB) Blanket are the ref... more The Helium Cooled Lithium Lead (HCLL) and the Helium Cooled Pebble Bed (HCPB) Blanket are the reference concepts in the European Breeding Blanket Programme for the DEMO design and for the related long term R&D. Recently, a similar design for both concepts has been developed, in particular both concepts use helium coolant and RAFM steel EUROFER as structural material. In
In the European Union, the basic manufacturing technologies have been selected for the two refere... more In the European Union, the basic manufacturing technologies have been selected for the two reference DEMO breeding blanket concepts: the helium cooled pebble bed and the helium cooled lithium lead. These technologies have been tested in the past years on small-scale samples and mock-ups, are presently under further development, and are now being adapted to the evolution of the two
The in-vessel components of the WENDELSTEIN 7-X stellarator consist of the divertor components an... more The in-vessel components of the WENDELSTEIN 7-X stellarator consist of the divertor components and the wall protection with its internal cooling supply. The main components of the open divertor are the vertical and horizontal target plates which form the pumping gap, the cryo-vacuum pumps and the control coils. The divertor volume is closed by graphite shielded baffle modules and with divertor closures. All these components are designed to be actively water-cooled. For the first commissioning phase planned in 2014, an inertial-cooled test divertor will be installed instead of the actively water-cooled high heat flux divertor. The wall protection consists of graphite-protected heat shields in the higher loaded areas and stainless steel panels in the lower loaded regions. The wall protection cooling circuits are connected through 80 supply-ports via so-called "plug-ins". It is envisaged to protect the diagnostic ports by paneltype port-liners. Special graphite-shielded port liners are used on the diagnostic injector and the neutral beam injector ports. The in-vessel components are mainly manufactured and tested at the Max-Planck-Institute für Plasmaphysik in its Garching workshop. Panels, high heat flux target elements and control coils are delivered by industrial partners. Manufacturing of the KiP ("Komponenten im Plasmagefäß") is in plan. Delivery of the components will be in time.
This article in an outline of the R&D progress on low Z materials and plasma facing components fo... more This article in an outline of the R&D progress on low Z materials and plasma facing components for NET. It covers the following topics: (1) neutron irradiation effects in low Z materials; (2) the evaluation of effects on plasma facing materials from off-normal operation events, i.e. disruption heat loads and runaway-electron events; (3) manufacture of first wall mock-ups with and without carbon protection tiles and simulation of the cyclic heat loads during normal operation; (4) development of design options for the divertor and a stepwise approach from materials screening tests to medium size mock-up tests for the design and development of the divertor.
... for 100 preparatory cycles and 900 cycles of 30 s heating and 30 s cooling time. At the end o... more ... for 100 preparatory cycles and 900 cycles of 30 s heating and 30 s cooling time. At the end of the 1000 cycles, the surface and the Be/DS-Cu joint of the mock up did not show any damage due to the fatigue test. Article Outline. ...
ABSTRACT In order to improve the tritium release capability of beryllium in helium cooled ceramic... more ABSTRACT In order to improve the tritium release capability of beryllium in helium cooled ceramic breeder (HCPB) type blankets, beryllium tablets with small grain sizes ranging from 0.4 to 20 μm and 15 mm diameter and a few millimetres height have been produced by a wet ball-milling process with subsequent hot pressing. Depending on process parameters, about 0.4–1.5 kg Be powder can be produced per day. X-ray investigations, SEM and EDX confirmed among others that the impurity concentrations could be kept very close to the raw material. As grain coarsening of Be is at a given temperature besides pressure and possible grain surface layers a sensitive function of grain size, Be tablets have been subjected to isochronal annealing for 2 h between 500 and 1000 °C in high vacuum. While annealing up to 800 °C did not change the average grain size of about 5 μm on the basis of the initial datasets, grain coarsening occurs clearly during 900 °C annealing.
One of the key issues of the European Helium Cooled Pebble Bed blanket is the behaviour under irr... more One of the key issues of the European Helium Cooled Pebble Bed blanket is the behaviour under irradiation of beryllium pebbles, which have the function of neutron multiplier. An intense production of helium occurs in-pile, as well as a non negligible generation of tritium. Helium bubbles induce swelling and a high tritium inventory is a safety issue. Extensive studies for a better understanding, characterisation and modelling of the behaviour of helium and tritium in irradiated beryllium pebbles are being carried out, with the final aim to enable a reliable prediction of gas release and swelling in the full range of operating and accidental conditions of a Fusion Power Reactor. The general strategy consists in integrating studies on macroscopic phenomena (gas release) with the characterisation of corresponding microscopic diffusion phenomena (bubble kinetics) and the assessment of some fundamental diffusion parameter for the models (gas atomic diffusion coefficients). The present work gives a summary of the latest achievements in this context. By an inverse analysis of experimental out-of-pile gas release from weakly irradiated pebbles, coupled to the study of the characteristics of bubble population, it has been possible to assess the thermal diffusion coefficients of helium and tritium in and to improve and validate the classical model of gas precipitation into bubbles inside the grain. The improvement of the description of gas atomic diffusion and precipitation is the first step to enable a more reliable prediction of gas release.
Equipos Nucleares S.A. as one of the contractors for the Wendelstein 7-X Stellarator project is c... more Equipos Nucleares S.A. as one of the contractors for the Wendelstein 7-X Stellarator project is currently manufacturing the coil support structure (CSS), which is a stainless steel ring-shaped structure formed by five equal sectors (modules), each of them made up from two symmetric half-modules, everything joined together by way of bolts and super-bolts. This CSS works as the central support for both planar and non-planar coils. The complicate shape of the CSS allows the ports to pass through the structure and reach their place into the vacuum vessel.
The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Pla... more The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Plasma Facing Components (PFC) which focused on the following main tasks: (a) the definition of basic design concepts for the First Wall (FW) and Divertor Plates (DP), (b) the analysis of the performance and likely lifetime of these PFC designs including the identification of major critical issues, (c) the start of R&D work giving already first results, and the definition of the required further R&D program to support the contemplated ITER Engineering Design Activity (EDA).
The long term European materials R&D prog... more The long term European materials R&D programme is mainly focused on the development of the following DEMO relevant materials: structural (RAFM steel called EUROFER), plasma facing (W alloy), breeder and neutron multiplier (Pb–17Li, Li4SiO4, Li2TiO3 and Be). Materials like ODS and SiCf/SiC composite are also investigated for advanced fusion reactor concepts. In addition, the materials programme includes activities in nuclear
The operation of W7-X stellarator for pulse length up to 30min with 10MW input power requires a f... more The operation of W7-X stellarator for pulse length up to 30min with 10MW input power requires a full set of actively water-cooled plasma facing components. From the lower thermally loaded area of the wall protection system designed for an averaged load of 100kW/m2 to the higher loaded area of the divertor up to 10MW/m2, various design and technological solutions have
The electrodes for the Wendelstein 7-X glow discharge system have been designed, tested and manuf... more The electrodes for the Wendelstein 7-X glow discharge system have been designed, tested and manufactured. The compact design relies on a cooled housing, integrated into the first wall cooling system, and a calotte-shaped graphite anode. The new mounting concept avoids the need of active cooling of the anode due to an improved thermal conduction. Comprehensive tests of a prototype electrode
All in vessel components (IVCs) of W7-X are actively cooled. Inside the plasma vessel about 4km o... more All in vessel components (IVCs) of W7-X are actively cooled. Inside the plasma vessel about 4km of pipes will be installed, supplying water to the IVC. 226 cooling circuits with 78 variants are necessary. The cooling circuits enter the cryostat and the plasma vessel through ad hoc flanged penetrations called “plug-ins”, which provide for the vacuum boundary between the plasma
Significant improvements of the ITER blanket were possible with the better definition of the ther... more Significant improvements of the ITER blanket were possible with the better definition of the thermal and electromagnetic effects. A supporting double wall backplate with reinforcement about the ports has been defined. The module has been simplified and the intrusion of the connections reduced. The attachment has been assessed by system analyses. The cooling system is configured to simplify the leak detection. 0-7803-4226-7/98/$10.00 0 1998 IEEE
The ITER General Design Requirements call for a design allowing the in-situ repair of protective ... more The ITER General Design Requirements call for a design allowing the in-situ repair of protective armours and, if possible, the separate replacement of FW sub-components, in order to minimise intervention time and radioactive waste generation. The technologies necessary to implement this maintenance strategy and the relevant experimental results of the on going ITER R&D program specifically tailored for their development are discussed in this paper .
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