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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.

In ITER [1] , it is foreseen to have two Test Blanket Systems (TBS) installed in two dedicated Equatorial Ports (EP) for the purpose of testing and validating different concepts of tritium breeding blankets [2] . Each TBS consists of a Test Blanket Module Port Plug (TBM PP) and its associated systems; the Pipe Forest (PF) in the Port Interspace (PI) area and the corresponding systems in the Port Cell (PC) area. While analyses have been conducted before [3] , the design has since gone through major evolution with a redesign of the PF, a new BP based on the butterfly doors concept, and an optional activated water delay tank. In this paper, the results of a comprehensive nuclear analyses that aim at assessing the neutronics performance of the revised PF and BP, will be discussed. These include calculations of neutron flux distributions and Shutdown Dose Rate (SDDR), following SA2 [4] scenario, in a baseline configuration and exploring the option of adding additional shielding plates in the PF. It has been found that the additional shielding plates reduces the neutron flux in the PI by ∼40–60% and the SDDR in the PI, due to activated components, by ∼50%.

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.

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.

This paper describes the main acceptance criteria and required acceptance tests for the components of the six Test Blanket Systems to be installed and operated in ITER. It summarizes the guide-line toward the establishment of detailed test plans for the TBS, starting from the end-product at the ITER Members factories, and to generally define the type of tests that have to be performed on the ITER site after shipment, and/or prior to the systems final commissioning phase.

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.

The irradiation response of PH13-8Mo stainless steel was measured up to 2.5 dpa at 200 and 300 °C irradiation temperatures. The PH13-8Mo, a martensitic precipitation-hardened steel, was produced by Hot Isostatic Pressing at 1030 °C. The fatigue tests (high cycle fatigue and fatigue crack propagation) showed a test temperature dependency but no irradiation effects. Tensile tests showed irradiation hardening (yield stress increase) of approximately 37% for 200 °C irradiated material tested at 60 °C and approximately 32% for 300 °C irradiated material tested at 60 °C. This contradicts the shift in reference temperature ( T 0 ) measured in toughness tests (Master Curve approach), where the Δ T 0 for 300 °C irradiated is approximately 170 °C and the Δ T 0 for the 200 °C irradiated is approximately 160 °C. This means that the irradiation hardening of PH13-8Mo steel is not suitable to predict the shift in the reference temperature for the Master Curve approach.

The irradiation-induced stress relaxation behavior of Eurofer97 at 300 °C up to 3.4 dpa 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 0.1 dpa, and after an irradiation dose of 2.7 dpa 42–47% of the original pre-stress is retained.

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.

In 2001 the Nuclear Research and Consultancy Group started a large graphite irradiation program for the development of High Temperature Reactor technology in the European framework. The irradiation experiments, containing present day available graphite grades, are performed at the High Flux Reactor in Petten. The grades are NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, LPEB, IG-110 and IG-430. In the fifth framework programme (2001–2004) and sixth framework programme (2005–2009) four irradiation experiments are foreseen, resulting in design curves at irradiation temperatures between 650 °C and 950 °C. The post-irradiation testing is focused on dimensional changes, dynamic Young’s modulus, coefficient of thermal expansion and coefficient of thermal conductivity. The irradiation programme and preliminary results from the first irradiation experiment at 750 °C to 8 dpa will be discussed in this paper.

Samples of Gilsocarbon graphite have been recently irradiated in the High Flux Reactor at Petten. They were irradiated in an inert atmosphere to a fast neutron dose of between 3 and 10 dpa, at temperatures of between 650 and 750 °C, this irradiation dose equates to approximately dimensional change ‘turnaround’. The present understanding of the irradiation induced microstructural changes considered to lead to the irradiation induced dimensional and material property changes are discussed and correlations between the various property changes noted. However, at present there are few microstructural studies to confirm or refute these theories. A methodology proposed to convert graphite CTE measurement, for all types graphite, from one temperature range to another is shown to be applicable to both unirradiated and irradiated Gilsocarbon graphite. In addition an important relationship for the thermal conductivity temperature dependence of irradiated Gilsocarbon graphite is shown to be reasonable.

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.

The irradiation programme EXOTIC (extraction of tritium in ceramics) is carried out within the European framework for the development of the helium cooled pebble bed concept. The EXOTIC-9/1 is the latest experiment in the series of EXOTICs that are irradiated in the high flux reactor in Petten. Tritium release and inventory in lithium containing ceramic pebbles are key properties to be tested in a TBM. New production routes of pebbles are developed, leading to different thermomechanical and tritium release properties. The objective of the EXOTIC-9/1 is to study in-pile tritium release behaviour of the latest developed lithiummetatitanate pebbles (Li2TiO3). The pebbles are produced by a extrusion–spheroidisation–sintering process at CEA. The new pebbles differ with respect to porosity from the lithiummetatitanate ceramics tested in the previous EXOTIC 8 programme. The pebbles have diameter in the range from 0.6 to 0.8 mm. Irradiation of EXOTC-9/1 started at 24 March 2005, and will continue until the end of 2006, in total about 400 irradiation days. The temperature is varied between 340 and 580 °C. Begin of Life (BOL) tritium production rate is 0.56 mCi/min. Based upon the in-pile tritium release measurements and the analysis of the tritium residence time it can be concluded that tritium release in the new batch of the high density Li2TiO3 pebbles irradiated in EXOTIC 9/1 is rather slow compared to the ceramics irradiated in the EXOTIC 8 irradiation campaign. In this paper, the in-pile tritium behaviour will be reported during normal operation and during transients in temperature, purge gas chemistry and gasflow. The collected data is compared to tritium release data from ceramics irradiated in previous EXOTIC experiments with respect to tritium inventory, residence time and porosity.

The main line of ceramic breeder materials research and development is based on the use of the breeder material in the form of pebble beds. At present, there are three candidate pebble materials (Li4SiO4, and two forms of Li2TiO3) for DEMO reactors that will be used for testing in ITER. This paper reviews the R&D of as-fabricated pebble materials against the blanket performance requirements and makes recommendations on necessary steps toward the qualification of these materials for testing in ITER.

This paper reports the objectives of a high dose irradiation programme of beryllium in the high flux reactor in Petten that will be performed in the frame of the European Programme for development of the helium cooled pebble bed (HCPB). The main objectives of irradiation are to study the irradiation behaviour of beryllium pebble beds, in terms of dimensional stability and their tritium retention behaviour. High fluence irradiation of beryllium specimens (HIDOBE) will be performed up to He contents of 6000 appm, which is about 30% of the expected DEMO end-of-life value with a relevant dPa to He ratio. In this paper, the nuclear parameters, irradiation conditions and the test-matrix, i.e. beryllium grades and pebbles are presented.

A high fluence irradiation of SiCf/SiC composites has been performed to study the effect of neutron irradiation on the mechanical and physical behaviour of those composites. The fibre reinforced silicon carbide composites have been irradiated at two temperature levels of 600 °C and 900 °C up to a fluence of 3.5 × 1025 n/m2 (E > 0.1 MeV). The stiffness of the bending bars after irradiation changed with factors between 0.75 and 1.04 for the different composites while the bending strength after irradiation was reduced with a factor of 2 in some cases. The laser flash thermal diffusivity ratio's (αirr/αo) of the composites measured at 600 °C are from 0.1 to 0.5 and from 0.25 to 0.75 for an irradiation temperature of 600 °C and 900 °C, respectively. The dimensional changes observed are small.

An overview is given of the wide range of activities contained within recent R&D being performed within Europe on In-vessel materials for structural, heat sink and plasma facing purposes for ITER. The effect of creep-fatigue interaction on the fatigue life of CuCrZr and the effects of irradiation on over-aged CuCrZr are given. In addition the lifetime of ITER components has been further investigated by the performance of in situ experiments with both neutrons and high-energy protons. The effect of hydrogen on the crack initiation fracture toughness of Ti is reported at a range of irradiation temperatures. The irradiation induced stress relaxation of Alloy 718 used for bolting applications is being studied and initial results will be described. Work in the area of plasma facing materials and re-welding issues will also be presented.

Since 1997, beryllide application as the neutron multiplier has been proposed in Japan and preliminary investigations revealed that some beryllides such as Be 12 Ti have excellent properties compared with beryllium metal. Recently, several international R&D programs were performed. Some pebbles of modified Be 12 Ti that included the Be phase were obtained by the rotating electrode method. Basic characterization using HIPped Be 12 Ti has also been performed. Preliminary experiments indicate that compatibility with stainless steel and tritium breeder materials, deuterium release, natural oxidation and oxidations by steam interaction were better than the behavior of beryllium. Preliminary neutron irradiation effects were studied in JMTR and by charged particle irradiation. It was observed that Be 12 Ti has less radiation damage than beryllium. In a future experiment, Japanese Be 12 Ti samples will be included in a long European irradiation program on beryllium to produce up to 6000 appm He. This is an IEA collaborative experiment, to evaluate neutron irradiation effects.

Lithium meta titanate pebbles has been prepared from agglomeration-sintering powders which were obtained by Li-Ti-peroxo-complex solution precursor (Li2TiO3 dissolved at room temperature in H2O + 40% H2O2 and stabilized with citric acid). Through this wet route Li2TiO3 pebbles with high density(∼92% of T.D.) has been obtained and the tritium release behavior has been tested `in-pile' by the EXOTIC-8.9 experiment (∼440 days of irradiation at full power in the high neutron flux of HFR-Petten). Tritium residence times (τ) in the pebbles has been measured during irradiation between 550 and 400 °C and He + 0.1%H2 purge gas composition. By a thermally activated process (activation energy=111 kJ/mol) with 410 °C as minimum temperature the tritium residence time is found to be about 1 day, which places this specimen in a good ranking position among those tested by the EXOTIC-series. A clear increase of the tritium release rate has been observed by increased H2 concentration (up to 1%) in the He purge. Out-of-pile ramp-annealing tritium desorption (TPD) tests on short-time irradiated pebbles has been also performed by various devices and conditions. The kinetic parameters from the TPD investigation gave consistent results with those characterizing the equilibrium times of tritium release rate after the gas composition and temperature transients imposed on the specimen during the in-pile experiment.

This paper reports on the pre-irradiation tests of SiC f /SiC materials as a contribution to the EFDA Technology Programme on advanced materials. A high temperature irradiation of SiC materials in the High flux Reactor in Petten, denoted as ‘SICCROWD’ has recently started. The SiC f /SiC materials included are supplied by the EU partners. Further materials are included in the frame of an IEA collaborative effort, by Japanese and USA partners. The irradiation matrix comprises a variety of fibres and matrix materials, as well as composites with 1-D, 2-D or 3-D architectures. CVD monolithic SiC serves as a reference material and should enable comparison with other, e.g. complementary, irradiation data. The specimen types are bars, for four-point bend tests, and discs for flash diffusivity measurements. Both types will be used to measure dimensional changes. Some specimens have been mounted to allow for in-situ determination of thermal conductivity degradation by neutron irradiation. The irradiation rig, materials and test matrix will be presented as well as the results of the pre-irradiation characterisations.

This paper reports on the thermomechanical behaviour of small pebble beds or pebble stacks in order to lay out irradiation test specimens. Uniaxial compression tests (UCTs) of the pebble stacks with relatively small diameters are performed at 27 and 800 °C i.e. the maximum irradiation temperature. The H/D ratio of the specimens influences the thermomechanical behaviour because of the friction between the pebbles and the steel tube. The results are compared with the thermomechanical data obtained from standard UCTs, which apply for larger pebble-beds. The outcome of the comparison will be used for the detailed design of the constrained pebble-beds in the high fluence irradiation. In addition, the compatibility between the ceramic breeder materials and the containment is studied. It has been shown that platinum has the best compatibility at 800 °C for both the lithium meta-titanate and lithium ortho-silicate ceramics.

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.

In the framework of developing the helium cooled pebble-bed (HCPB) blanket an irradiation test of pebble-bed assemblies is prepared at the HFR Petten. The test objective is to concentrate on the effect of neutron irradiation on the thermal-mechanical behaviour of the HCPB breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. The paper reports on the project status, and presents the results of pre-tests, material characteristics, the manufacturing of the pebble-bed assemblies, and the nuclear and thermo-mechanical loading parameters.

In the frame of the ITER task BL 14.1, the residual strength of precipitation hardened Inconel 718 bolts, preloaded to stresses up to 90% of the yield strength, after neutron irradiation has been investigated. In the ITER design, the Inconel 718 bolts are used to attach the backplate to the radial supports and may experience large shocks from off-normal events during operation. Three irradiation rigs, containing 9 bolts each, were irradiated at the HFR at dose levels of 0.2, 0.7 and 1.1 dpa, respectively, at a temperature of 573 K. The bolts were designed to resemble a scaled-down version of the actual design and the irradiation conditions were chosen to resemble the actual operation conditions. Our results show that two competitive processes occur during neutron irradiation. At low dose, dissolution of precipitates is dominant, which causes a softening of the material. At higher doses, irradiation hardening takes place. The combined effect of the two processes is that the irradiation hardening is relatively small compared to other materials.

Li2ZrO3 and Li2TiO3 pebbles are being investigated at Commissariat a 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.

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 Li 2 TiO 3 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 Li 4 SiO 4 pebble bed is measured along with the tritium permeation rate through a T91 tube with a Fe–Al/Al 2 O 3 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.

The world-wide efforts on ceramic breeder materials in the last two years concerned Li 2 O, Li 4 SiO 4 , Li 2 TiO 3 and Li 2 ZrO 3 , with a clear emphasis on the development of Li 2 TiO 3 . 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 (RD tritium release and retention properties; determination of the key factors limiting blanket life.

An in-pile test of helium cooled pebble-bed (HCPB) typical pebble-bed assemblies will be performed in the high flux reactor (HFR) Petten. Its major objective is to study the thermo-mechanical behaviour of the ceramic breeder pebble-beds and neutron irradiation effects. In order to arrive at the final design of test elements, results have been obtained from out-of-pile pre-tests and modelling activities. Preliminary thermo-mechanical analyses of the behaviour of the in-pile test-elements have been performed.

Lithium metatitanate (Li2TiO3) is considered as one of the candidate materials for the ceramic breeder in both the ITER Breeding Blanket and the European Helium Cooled Pebble-Bed Blanket for DEMO. A wet process based on powder-gelation for the manufacture of Li2TiO3-pebbles is described. Pebble characteristics and its basic properties are given along with results of out-of-pile tritium release experiments.

In the framework of the European blanket development programme a number of ceramic breeder materials have been considered for the DEMO blanket concepts under development in the period 1992–1994. Candidate materials have been irradiated in the HFR Petten in a series of experiments codenamed EXOTIC. The objective of the latest series, EXOTIC-7, was to reach target and peak lithium burn-ups, representative for so-called DEMO end-of-life conditions. Further objectives were to investigate the burn-up effects on the mechanical stability and on the tritium release characteristics of the candidate ceramic breeder materials. In pile performance and results of the post irradiation examinations are discussed for metazirconates and orthosilicates, materials now considered as candidates for both ITER and DEMO blankets.

The EXOTIC-7 irradiation experiment in the high flux reactor (HFR) has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characteristics of candidate ceramic breeder materials, originating from the fusion programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li2ZrO3, LiAlO2 and Li8ZrO6 and pebbles of Li4SiO4 and Li2ZrO3, with a variety of characteristics, like grain size and porosity. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules contained a mixture of Li4SiO4 and beryllium pebbles. The experimental design, sample loading and main irradiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented.

The long-term objective of the European Fusion Development Agreement (EFDA) fusion materials programme is to develop structural and armor materials in combination with the necessary production and fabrication technologies for reactor concepts beyond the International Thermonuclear Experimental Reactor. The programmatic roadmap is structured into four engineering research lines which comprise fabrication process development, structural material development, armor material optimization, and irradiation performance testing, which are complemented by a fundamental research programme on "Materials Science and Modeling." This paper presents the current research status of the EFDA experimental and testing investigations, and gives a detailed overview of the latest results on materials research, fabrication, joining, high heat flux testing, plasticity studies, modeling, and validation experiments. (C) 2013 Elsevier B.V. All rights reserved.

Carbon base candidate plasma facing materials have been subjected in the enhanced high power ECN-laser facility to short pulse high heat loads as anticipated to occur during plasma disruptions in next step devices like the International Thermonuclear Experimental Reactor or the Next European Torus. Recently, the first results of measurements on transient surface temperature response up to 3000 K, below erosion threshold, have been presented. After additional modification of the pyrometer, temperatures up to 4000 K can now be measured, permitting analysis of surface temperature response in relation to erosion. Qualitatively, the measured temperature responses are very similar to numerical calculations, and show clearly if erosion is taking place. Quantitatively the interpretation requires some caution, but reasonable fits are obtained. Some results of multiple shot experiments on carbon base materials are also presented. The incremental erosion is found to decrease only slighly after the first shot.

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.

Plasma-sprayed boron carbide coatings have been manufactured by different suppliers onto substrates of type 316L stainless steel. The coating thickness ranges from 0.3 to 2.0 mm. The larger thicknesses could only be achieved by application of an adaptive or gradient bond-layer between substrate and the boron carbide top coating. Measurements of the thermal diffusivity of coating materials are reported. Several high heat flux facilities have been used to study the thermal shock and erosion behaviour of the coated samples. A supporting numerical analysis of the thermal behaviour of the coating under normal and off-normal heat loads is presented, focussing on the differences between electron beam and laser beam tests due to volumetric energy deposition. Some aspects of the applicability of plasma sprayed B 4 C coatings for first-wall protection in a next step device are discussed.

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 technologies for future divertor concepts. The programmatic roadmap is structured into four engineering research lines which comprise fabrication process development, structural material development, armor material optimization, and irradiation performance testing, which are complemented by a fundamental research programme on "Materials Science and Modeling". This paper presents the current research status of the EFDA experimental and testing investigations, and gives a detailed overview of the latest results on fabrication, joining, high heat flux testing, plasticity, modeling, and validation experiments. (C) 2011 Elsevier B.V. All rights reserved.

The mechanical properties of two explosively formed saddle shaped 60 mm thick plates of 316L(N)-IG steel with and without an electron beam weld have been investigated. Two different conditions have been characterized: (1) Reference condition and (2) ITER relevant condition. The reference material exhibits consistent results for both plates, mechanical properties of reference material are similar to the properties previously observed for 316L(N)-IG steels. No significant difference in mechanical properties and microstructure between different positions in the 60 mm plate is observed. Tensile properties for ITER relevant materials are found to comply both with the RCC-MR code qualified data for 316L(N) steel used for the structural design and with ITER Materials Properties Handbook. As expected total elongation and uniform elongation for weld material are lower than the average curves obtained for the base material. © 2010 Elsevier B.V. All rights reserved.

An important issue for the next step thermonuclear fusion devices is the development of a reliable engineering solution for the first wall and in particular for the divertor. Besides severe mechanical loads, these plasma facing components (PFC) will be subjected to energetic pulses during plasma disruptions. To evaluate the performance of plasma facing materials and thus to predict the lifetime of the PFCs, simulation experiments have been performed in the JAERI electron beam irradiation stand (JEBIS). Here different candidate plasma facing materials were evaluated in short beam pulses (1.2–10 ms) with an energy deposition of 2–9 MJ m −2 . The response of the individual materials to single and multiple shot beam pulses was investigated, and in these analyses special attention was given to the material erosion (melting, sublimation, particle emission). The quantification of these effects was done by weight loss measurements and by optical profilometry on the ablation craters. In addition, microstructural and morphological changes in the loaded surface were investigated.

New experimental results from disruption heat flux simulations in the millisecond range with laser and electron beams are discussed. For a number of graphites, boronated graphites and carbon fiber composites, the effective enthalpy of ablation is determined as 30±3 MJ/kg, using laser pulses of about 0.3 ms. No effect of boron doping on the ablation enthalpy is found. A simple parametric study of vapour shielding is presented, with shield transmission factors from 5 to 100% and compared with recently published results from pulsed plasma devices in Russia. Estimates for the disruption erosion lifetime of plasma facing components in ITER are briefly discussed.

An evaluation is given for the prediction for disruption erosion in the International Thermonuclear Engineering Reactor (ITER). At first, a description is given of the relation between plasma operating parameters and system dimensions to the predictions of loading parameters of Plasma Facing Components (PFC) in off-normal events. Numerical results from ITER parties on the prediction of disruption erosion are compared for a few typical cases and discussed. Apart from some differences in the codes, the observed discrepancies can be ascribed to different input data of material properties and boundary conditions. Some physical models for vapour shielding and their effects on numerical results are mentioned. Experimental results from ITER parties, obtained with electron and laser beams, are also compared. Erosion rates for the candidate ITER PFC materials are shown to depend very strongly on the energy deposition parameters, which are based on plasma physics considerations, and on the assumed material loss mechanisms. Lifetimes estimates for divertor plate and first wall armour are given for carbon, tungsten and beryllium, based on the erosion in thermal quench phase.

The effects of off-normal heat loads accompanying plasmadisruptions have been studied by both experimental and numerical simulations. Experiments have been performed on a range of polygranular graphites, pyrolytic graphite and carbon composite material. The measured erosion is compared with numerical predictions by a transient heat load code taking a revised value for the vaporization enthalpy of carbon. The effect of variations in thermo-physical material parameters on thermal erosion behaviour is discussed.

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.

We present a patient with anaemia and an increased erythrocyte sedimentation rate in which the diagnosis of a T3 renal cell carcinoma with a level IV tumour thrombus was made with fusion of CT and [F-18]-fluorodeoxyglucose (FDG)-positron emission tomography (PET) imaging. The tumour thrombus would have been missed just by using CT scanning prior to the operation. Only by means of FDG-PET imaging the suspicion of a thrombus was made and this was confirmed by transoesophageal ultrasound. CT with a scan delay of 60 s was chosen because CT indication was anaemia with elevated erythrocyte sedimentation rate. This CT protocol may not be optimal for good opacification of the vena cava inferior. Determination of the true level of tumour thrombus is of utmost importance for planning the operation.

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.

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.

The first wall in NET is made of alloys which melt locally, when hit by plasma disruptions or run away electrons. Therefore several ceramic and carbon materials are candidates as protection coatings or tiles. Results of disruption simulation with a multi-mode Nd: YAG pulsed laser are reported. Pulse durations are in the range of 0.2–20 ms and energy densities are up to 10 MJ/m2. These quantities are in line with predictions for NET disruptions. Austenitic stainless steel AISI 316, a graphite material and titanium carbide, materials relevant to the NET design, are studied. The phenomena due to the laser energy deposition are described and quantified in terms of melting depth and erosion loss. These quantities are related to predictions obtained by numerical calculations.

The current magnetic confinement nuclear fusion power reactor concepts going beyond ITER are based on assumptions about the availability of materials with extreme mechanical, heat, and neutron load capacity. In Europe, the development of such structural and armour materials together with the necessary production, machining, and fabrication technologies is pursued within the EFDA long-term fusion materials programme. This paper reviews the progress of work within the programme in the area of tungsten and tungsten alloys. Results, conclusions, and future projections are summarized for each of the programme's main subtopics, which are: (1) fabrication, (2) structural W materials, (3) W armour materials, and (4) materials science and modelling. It gives a detailed overview of the latest results on materials research, fabrication processes, joining options, high heat flux testing, plasticity studies, modelling, and validation experiments. (C) 2012 Elsevier B.V. All rights reserved.