- Senior officer on nuclear technologies for energyedit
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2012.08.018 This is a PDF file of an unedited manuscript that has been accepted for publication. As a service to our customers
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The difficulties associated with beam profile measurements of pulsed high power lasers have been recognized but they have not been solved completely. This is especially the case for high power Nd:YAG lasers with repetition rates up to a... more
The difficulties associated with beam profile measurements of pulsed high power lasers have been recognized but they have not been solved completely. This is especially the case for high power Nd:YAG lasers with repetition rates up to a few 100 Hz. Most diagnostic devices have a limited application window due to a fixed scanning frequency, a high minimum scan duration or a limited allowable laser beam power. The device presented here is based on the concept of the rotating needle and is capable of determining the beam profiles of commonly used flashlamp excited solid state lasers. The paper describes the working principle and lay-out of the ECNUTRON laser beam analyzer. The device has a compact detection head, variable scan frequency up to 400 Hz, variable beam diameter up to 15mm and a minimum scan duration of 10-4s. The spectral response is flat for the VIS and NIR spectral region. The pulse synchronisation and trigger delay allow for a time dependent profile measurement by scanning subsequent pulses. The beam profile consists of two orthogonal cross sections which can be displayed in real time. The spatial resolution is about 5–103 μm2. Experimental results with a Nd:YAG laser are given.
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Fusion power plant operation will strongly depend on the economy and reliability of crucial components, such as first wall modules, tritium breeding blankets and divertors. Their operating temperature shall be high to accomplish high... more
Fusion power plant operation will strongly depend on the economy and reliability of crucial components, such as first wall modules, tritium breeding blankets and divertors. Their operating temperature shall be high to accomplish high plant efficiency. The materials properties and component fabrication routes shall also assure long reliable operation to minimize plant outage. The components must be fabricated in large quantities based on demonstrations with a limited amount of test beds. Mock-ups and test loops will, through iteration processes, demonstrate the reliable operation under reference thermal-hydraulic conditions. 14 MeV neutrons escaping the plasma dominate the nuclear conditions near the first wall. Neutron transport analyses have shown that large portions of the components near the plasma have to cope with a neutron spectrum resembling that of a fission core. MTR's, Materials Test Reactors, have been used successfully to test components for fusion power plants. Thei...
This paper describes the newly designed irradiation rigs capable of testing the mechanical stability of DWT's and the performance of newly developed tritium permeation barriers of the WCLL blanket under realistic thermal-hydraulic and... more
This paper describes the newly designed irradiation rigs capable of testing the mechanical stability of DWT's and the performance of newly developed tritium permeation barriers of the WCLL blanket under realistic thermal-hydraulic and neutronic conditions at the HFR Petten.
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Lithium titanate is one of the candidate materials for tritium breeding blankets in fusion devices. The Helium Cooled Pebble-Bed (HCPB) blanket concept foresees the breeder ceramic shaped as pebbles with typical diameters in the range of... more
Lithium titanate is one of the candidate materials for tritium breeding blankets in fusion devices. The Helium Cooled Pebble-Bed (HCPB) blanket concept foresees the breeder ceramic shaped as pebbles with typical diameters in the range of 0.2 to 1.5 mm. Ceramics presently considered as candidates are Li 4 SiO 4 , Li 2 TiO 3 and Li 2 ZrO 3 . In the framework of the European Blanket Project several lithium ceramics are being irradiated in the High Flux Reactor (HFR) in Petten, with in-situ extraction of tritium. Results from three experiments on lithium titanate from the EXOTIC-8 series are reported.
Abstract The ITER project aims to demonstrate scientific and technological feasibility of controlled energy production through thermonuclear fusion and the IAEA plays a pro-active role in catalyzing innovation and enhancing the worldwide... more
Abstract The ITER project aims to demonstrate scientific and technological feasibility of controlled energy production through thermonuclear fusion and the IAEA plays a pro-active role in catalyzing innovation and enhancing the worldwide commitment to fusion. The 3rd Joint IAEA-ITER Technical Meeting on ITER Materials & Technologies, held in November 2015 has been aimed to contribute to the development of a knowledge base of properties, processes and technologies relevant to ITER structural and plasma-facing materials/components. This paper is a summary of the presentations and discussions of the meeting, which was mainly devoted to materials and technologies for ITER, while those relevant beyond, like a DEMO were considered as well within the scope.
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ABSTRACT Irradiation experiments are performed in support of fusion blanket technology development. These comprise ceramic solid breeder materials, and a liquid Lithium Lead alloy, as well as blanket subassemblies and components.... more
ABSTRACT Irradiation experiments are performed in support of fusion blanket technology development. These comprise ceramic solid breeder materials, and a liquid Lithium Lead alloy, as well as blanket subassemblies and components. Experimental facilities at the HFR to study tritium release, permeation characteristics, and neutron irradiation performance, have recently been extended. This paper gives an overview on the tritium breeding materials irradiation programme and describes the facilities required for irradiation testing and on-line tritium measurement.
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Abstract Test Blanket Systems (TBS) will be operated in ITER in order to prepare the next steps towards fusion power generation. After the initial operation in H/He plasmas, the introduction of D and T in ITER will mark the transition to... more
Abstract Test Blanket Systems (TBS) will be operated in ITER in order to prepare the next steps towards fusion power generation. After the initial operation in H/He plasmas, the introduction of D and T in ITER will mark the transition to nuclear operation. The significant fusion neutron production will give rise to nuclear heating and tritium breeding in the in-vessel part of the TBS. The management of the activated and tritiated structures of the TBS from operation in ITER is described. The TBS specific features like tritium breeding and power conversion at elevated temperatures, and the use of novel materials require a dedicated approach, which could be different to that needed for the other ITER equipment.
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Abstract The paper describes the organization of the Test Blanket Module (TBM) program, its overall objective and schedule and the status of the technical activities within the ITER Organization-Central Team (IO-CT). The latter include... more
Abstract The paper describes the organization of the Test Blanket Module (TBM) program, its overall objective and schedule and the status of the technical activities within the ITER Organization-Central Team (IO-CT). The latter include the design integration of the Test Blanket Systems (TBSs) into the nuclear buildings, ensuring all interfaces with other ITER systems, the design of the common TBS components such as the TBM Frames, the Dummy TBMs, and the TBS maintenance tools and equipment in the TBM Port Cell as well as in the Hot Cell building, the design of the TBS connection pipes and the definition of the required maintenance operations and associated R&D. The paper also discusses the major challenges that the TBM Program will be facing in ITER such as the potential impact of the TBMs ferritic/martensitic structures on plasma operations, the approaches to tritium and contamination confinement, the required mitigation and recovery actions in case of accidents, and the assessment of the reliability aspects that could have an impact on ITER availability.
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The plasma disruption heat load is simulated experimentally using a pulsed laser beam with high energy density and short pulse duration (0.2-20 ms) covering a certain range of ITER design values. The present status of the laser heat flux... more
The plasma disruption heat load is simulated experimentally using a pulsed laser beam with high energy density and short pulse duration (0.2-20 ms) covering a certain range of ITER design values. The present status of the laser heat flux test facility and new experimental tools are described. Spatial and time resolved profiles of the laser beam are given. Experimental results are presented including the variation of angle of incidence of the laser beam relative to the material surface. The nature and effects of the induced vapour plume are discussed. Materials studied are relevant to the ITER design. Experimental results are compared with numerical calculations. Some implications for the design of First Wall and Divertor of ITER are addressed.
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An alternative to the basic NET first wall (FW). concept is presented. The concept aims at: minimizing the maximum surface temperature of the protection material, the impurity influx from the wall, the intervention frequency for... more
An alternative to the basic NET first wall (FW). concept is presented. The concept aims at: minimizing the maximum surface temperature of the protection material, the impurity influx from the wall, the intervention frequency for maintenance as well as increasing the operating margins against the heat loads. This will allow to reduce: the required expansion volumes in case of accidental water ingress in the plasma chamber and the tritium inventory present in the protection material. The solution presented is made of removable local limiters (LL) and the proper first wall in between. The main role of the local limiters is to intercept the particle flux during burn (α particle), during start up when they act as plasma limiters and during disruption to stop the run away electrons. The technologies envisaged to make the local limiter are already being developed for the divertor or they could be a limited extrapolation of existing one. The FW will be coated with methods already existing (e.g. plasma spray or physical vapor deposition), but they should be qualified for the operating condition of the first wall and in-situ repair. The coating can be directly applied on the blanket box or on tubes which afterwards are assembled on the blanket box.
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A coated solution for First Wall (FW) protection in a next step device is being developed by the NET Team. The FW parts which receive only modest heat loads under normal and off-normal operation could be protected by a coating of low Z... more
A coated solution for First Wall (FW) protection in a next step device is being developed by the NET Team. The FW parts which receive only modest heat loads under normal and off-normal operation could be protected by a coating of low Z material. The potential use of plasma-sprayed boron carbide (B4C) coatings is discussed. A global thermal analysis of coating behaviour under normal and off-normal conditions is presented. Recent results from several high heat flux facilities on the thermal shock and erosion behaviour of plasma sprayed B4C coatings are discussed. These results indicate that coating development has progressed in a way that a First Wall protective coating has the potential to survive at least several tens and possibly hundreds of disruptions.
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First results of experimental investigations on the laser reweldability of neutron irradiated material are reported. These experiments include the manufacture of 'heterogeneous' joints, which means joining of irradiated stainless... more
First results of experimental investigations on the laser reweldability of neutron irradiated material are reported. These experiments include the manufacture of 'heterogeneous' joints, which means joining of irradiated stainless steel of type AISI 316L-SPH to 'fresh' unirradiated material. The newly developed laser welding facility in the ECN Hot Cell Laboratory and experimental procedures are described. Visual inspections of welded joints are reported as well as results of electron microscopy and preliminary metallographic examinations.
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This article reviews the research on ceramic tritium breeding materials and the development of these materials, which aims at providing the tritium required for the deuterium–tritium fuel cycle in magnetic fusion reactors. In particular,... more
This article reviews the research on ceramic tritium breeding materials and the development of these materials, which aims at providing the tritium required for the deuterium–tritium fuel cycle in magnetic fusion reactors. In particular, the performance of various oxides of lithium is discussed. Currently, much of the ceramic breeder research and development is focused on lithium orthosilicate and metatitanate systems in the form of pebble beds. This chapter reviews design requirements, manufacturing routes, pebble and pebble-bed thermomechanics, tritium production and release properties, neutron-irradiation behavior, chemistry, and modeling. Ongoing work is summarized and the outlook for further work, including test programs in International Thermonuclear Experimental Reactor (ITER), is provided.
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ABSTRACT Graphite is used as a moderator, reflector and structural component in pebble bed and prism High Temperature Reactors (HTRs). It is fortunate to reactor designers that irradiated graphite shows remarkably high creep behaviour... more
ABSTRACT Graphite is used as a moderator, reflector and structural component in pebble bed and prism High Temperature Reactors (HTRs). It is fortunate to reactor designers that irradiated graphite shows remarkably high creep behaviour under the influence of fast neutron irradiation at temperatures far below those required for significant creep strains to be generated in unirradiated graphite. This creep behaviour is important in the design of nuclear graphite reactor cores because the self-induced shrinkage stresses generated in typical core components during irradiation can be relieved. However, there are no reliable data on high fluence irradiation creep and the mechanistic understanding of the irradiation creep is insufficiently developed to reliably extrapolate to the high fluences expected of graphite in future HTR designs. The understanding of irradiation creep is further complicated because it has been experimentally observed that irradiation creep strain in graphite modifies other properties in particular the coefficient of thermal expansion. In addition modified changes in Young's modulus in crept specimens have been reported and it has also been postulated that irradiation creep may also modify dimensional change. The assessment of irradiation creep in graphite components is based on empirical laws derived from data obtained from small samples irradiated in a materials test reactor. However, due to the complicated irradiation rigs required and the amount of dimensional and property measurements needed to be taken, constant stress irradiation creep experiments are difficult and very expensive to carry out successfully. However, restrained creep experiments are simple to implement, less expensive and can be easily included as part of other, more conventional irradiation graphite experimental programmes. However, in the past, the disadvantage of these experiments has been that the results have been difficult to interpret using the then available analytical methods. In this paper the restrained creep experiment is revisited and analysed numerically and the possible benefit of using a restrained creep experiment in future graphite irradiation experiments is investigated. It is shown that a numerical simulation of the restrained creep experiment behaviour would be an essential tool to ensure that the stress within the specimen remains within defined limits so that specimen failure can be avoided.
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ABSTRACT In two recent irradiation experiments in the HFR Petten, tritium permeation rates through representative materials to be used as cooling tubes of the water-cooled lithium-lead blanket have been measured in-pile. These latest... more
ABSTRACT In two recent irradiation experiments in the HFR Petten, tritium permeation rates through representative materials to be used as cooling tubes of the water-cooled lithium-lead blanket have been measured in-pile. These latest experiments in the EXOTIC 8 series (E 8.9 and E 8.10) are made of a double wall tube (DWT) and a T91 tube with an Fe–Al/Al2O3 layer acting as tritium permeation barrier (TPB). These tubes contain annular pebble beds of ceramic breeder materials for the helium-cooled pebble bed concept blanket as tritium breeding material. Both experiments are built up of two concentric and independently purged containments allowing on-line tritium release rate and permeation rate measurements. In-pile operation has ended in March 2001 after 450 full power days and resulted in an irradiation damage of approximately 2.6 and 3.2 dpa, respectively in T91 steel. This paper reports on the experimental results obtained for in-pile tritium permeation and discusses the influence of purge gas compositions, temperature and irradiation on tritium permeation through the DWT and TPB.
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ABSTRACT Two new irradiation projects are being performed at the HFR Petten, named EXOTIC-8.9 and EXOTIC-8.10. Issues such as tritium release from candidate ceramic breeder pebbles for the HCPB blanket and tritium permeation through... more
ABSTRACT Two new irradiation projects are being performed at the HFR Petten, named EXOTIC-8.9 and EXOTIC-8.10. Issues such as tritium release from candidate ceramic breeder pebbles for the HCPB blanket and tritium permeation through cooling tubes of the WCLL blanket are investigated simultaneously. In EXOTIC-8.9, the tritium release behaviour of a Li2TiO3 pebble bed is measured along with the tritium-permeation rate through a double-wall tube (DWT) of T91 with a Cu interlayer. In EXOTIC-8.10, the tritium release behaviour of a Li4SiO4 pebble bed is measured along with the tritium permeation rate through a T91 tube with a Fe–Al/Al2O3 coating as tritium permeation barrier (TPB). Tritium permeation phenomena are studied by variations of temperatures and purge gas conditions. This paper reports on the results of the first 100 irradiation days.
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ABSTRACT Li2ZrO3 and Li2TiO3 pebbles are being investigated at Commissariat à l'Energie Atomique as candidate alternative ceramics for the European helium-cooled pebble bed (HCPB) blanket. The pebbles are fabricated using the... more
ABSTRACT Li2ZrO3 and Li2TiO3 pebbles are being investigated at Commissariat à l'Energie Atomique as candidate alternative ceramics for the European helium-cooled pebble bed (HCPB) blanket. The pebbles are fabricated using the extrusion–spheronization–sintering process and are optimized regarding composition, geometrical characteristics, microstructural characteristics, and material purity. Tests were designed and are being performed with other organizations so as to check the functional performance of the pebbles and pebble beds with respect to the HCPB blanket requirements, and, finally, to make the selection of the most appropriate ceramic for the HCPB blanket. Tests include high temperature long-term annealing, thermal shock, thermal cycling, thermal mechanical behaviour of pebble beds, thermal conductivity of pebble beds, and tritium extraction. Current results indicate the attractiveness of these ceramics pebbles for the HCPB blanket.
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ABSTRACT The world-wide efforts on ceramic breeder materials in the last two years concerned Li2O, Li4SiO4, Li2TiO3 and Li2ZrO3, with a clear emphasis on the development of Li2TiO3. Pebble-manufacturing processes have been developed up to... more
ABSTRACT The world-wide efforts on ceramic breeder materials in the last two years concerned Li2O, Li4SiO4, Li2TiO3 and Li2ZrO3, with a clear emphasis on the development of Li2TiO3. Pebble-manufacturing processes have been developed up to a 10 kg scale. Characterisation of materials has advanced. A jump-wise progress is observed in the characterisation of pebble-beds, in particular of their thermo-mechanical behaviour. Thermal property data are still limited. A number of breeder materials have been or are being irradiated in material test reactors like HFR and JMTR. The EXOTIC-8 series of in-pile experiments is a major source of tritium release data. This paper discusses the technical advancements and proposes a focus for further research and development (R&D) : pebble-bed mechanical and thermal behaviour and its interactions with the blanket structure as a function of temperature, burn-up, irradiation dose and time; tritium release and retention properties; determination of the key factors limiting blanket life.
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ABSTRACT The irradiation-induced stress relaxation behavior of Eurofer97 at 300°C up to 3.4dpa and under pre-stress loads typical for the ITER applications is investigated. The bolt specimens are pre-loaded from 30% to 90% of the yield... more
ABSTRACT The irradiation-induced stress relaxation behavior of Eurofer97 at 300°C up to 3.4dpa and under pre-stress loads typical for the ITER applications is investigated. The bolt specimens are pre-loaded from 30% to 90% of the yield strength. To verify the results obtained with the pre-stressed bolts, bent strips were investigated as well. The strips are bent into a pre-defined radius in order to achieve similar pre-stress levels. The irradiation-induced stress relaxation is found to be independent of the pre-stress level. 10–12% of the stress relaxation in Eurofer97 may be reached after a dose of 0.1dpa, and after an irradiation dose of 2.7dpa 42–47% of the original pre-stress is retained.
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ABSTRACT This paper is an overview of the very first results obtained on pure tungsten (W) and oxide dispersed strengthened (ODS) W alloys produced by the Metal Injection Molding (MIM) technique for fusion applications. An extensive... more
ABSTRACT This paper is an overview of the very first results obtained on pure tungsten (W) and oxide dispersed strengthened (ODS) W alloys produced by the Metal Injection Molding (MIM) technique for fusion applications. An extensive mechanical and physical characterization was performed, together with microstructural material investigation. The reported work was accomplished within the framework of the European Fusion Development Agreement work program. The main objective was to develop suitable tungsten grades for structural and armor divertor applications in the future DEMO fusion reactor.
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Abstract This study presents experimental disruption simulation results using a pulsed high power laser beam with high energy density (0–20 MJ/m 2 ) and short pulse duration (0.2–20 ms) under vacuum conditions. Time-dependent profiles of... more
Abstract This study presents experimental disruption simulation results using a pulsed high power laser beam with high energy density (0–20 MJ/m 2 ) and short pulse duration (0.2–20 ms) under vacuum conditions. Time-dependent profiles of the laser beam are given. The paper deals mainly with the European reference heat austenitic steel Type AISI 316L. The melting effects are quantified and compared with numerical predictions by a transient heat load code. The effects of repeated disruptions are analyzed. The extrapolation of experimental data obtained with small spots to predictions for larger areas is addressed.
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All the recent DEMO design studies for helium cooled divertors utilize tungsten materials and alloys, mainly due to their high temperature strength, good thermal conductivity, low erosion, and comparably low activation under neutron... more
All the recent DEMO design studies for helium cooled divertors utilize tungsten materials and alloys, mainly due to their high temperature strength, good thermal conductivity, low erosion, and comparably low activation under neutron irradiation. The long-term objective of the EFDA fusion materials programme is to develop structural as well as armor materials in combination with the necessary production and fabrication