Characterization of Plasma Sprayed Beryllium ITER First Wall Mockups (original) (raw)
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Physica Scripta, 2007
In order to perform a fully integrated material test, JET has launched the ITER-like wall project with the aim of installing a full metal wall during the next major shutdown. The material foreseen for the main chamber wall is bulk Be at the limiters and Be coatings on inconel tiles elsewhere. R&D process comprises global characterization (structure, purity etc) of the evaporated films and testing of their performance under heat loads. The major results are (i) the layers have survived energy loads of 20 MJ m −2 which is significantly above the required level of 5-10 MJ m −2 , (ii) melting limit of beryllium coating would be at the energy level of 30 MJ m −2 , (iii) cyclic thermal load of 10 MJ m −2 for up to 50 cycles have not induced any noticeable damage such as flaking or detachment.
The structure, properties and performance of plasma-sprayed beryllium for fusion applications
Physica Scripta, 1996
Plasma-spray technology is under investigation as a method for producing high thermal conductivity beryllium coatings for use in magnetic fusion applications. Recent investigations have focused on optimizing the plasma-spray process for depositing beryllium coatings on darnaged beryllium surfaces, Of particular interest has been optimizing the processing parameters to maximize the through-thickness thermal conductivity of the beryllium coatings. Experimental results will be reported on the use of secondary Hz gas additions to improve the melting of the beryllium powder and transferred-arc cleaning to improve the bonding between the beryllium coatings and the underlying surface. Information will also be presented on therrrwl fatigue tests which were done on beryllium coated ISX-B beryllium limiter tiles using 10 sec cycle times with 60 sec cooldowns and an International Thermonuclear Expcrimentrd Reactor (ITER) relevant divertor heat flux slightly in excess of 5 MW/m2. DMcLAIMER Thlt rq!oti ww proptird M m mwunt of work IporMod by an tgoncy of tho Unltcd MSIW Oovsrnmant, Ndther Ihs Unkod Statw Oovommant nor wry a~sncy thoroof, nor My of thdr cmployWkmnkw any wm?snty, wproM or Impll.d, or ustumm rnnylqal liabllity or Wont}. blllty for the wcurscy, complstonom, or uwlulncns of any Irrformstion, sppcrclus, produat, or proms dhcbtd, or rsprw$nts \hat III wc would not infrlngs pdvrntoly owrwd rights, Rafcr. once hd! to any rIpwific wmmorclul product,prowM, or SAW by irudo name, trudamtrk, m~nufdurcr, Ur othorwlM doa not nwwwrily Corwliluto of Imply Ill ondorwmantt r~m" mondatlon, or fivorlng by the Urrltod !ltstas (40\'crnmcnt or any @gcncy thormf, The vlaws u&.@ninnn of @hors Mpwsd hcroht do not nocadly ntato or r.fkt th~O(th. Urrltcd Ww Clovcrnmont or my cgonay Ihor@f.
Advanced solutions for beryllium and tungsten plasma-facing components
Fusion Engineering and Design, 1998
Beryllium and tungsten are candidate plasma-facing armour materials for the International Thermonuclear Experimental Reactor (ITER). These armours are proposed for areas with low heat flux ( 0 5 MW m − 2 ); however, in the divertor, surface melting during abnormal events may occur. This paper reports the progress made in developing novel approaches to solving the difficulties posed in designing with these armours. A Be monoblock brazed to an oxygen free high conductivity (OFHC) 10 mm ID Cu tube using InCuSil 'ABA' braze alloy has survived 130 cycles of 10-11 MW m − 2 for 6 s, with surface temperatures of 1250°C. No visible surface cracking occurred. The same monoblock was then exposed to several cycles of 20 -22 MW m − 2 for 8 s, creating a 2 mm deep molten layer. High cycle fatigue was then performed. The test results are detailed in this paper. Comparison between experimental and theoretical results are made. W and Cu have a large mismatch in their thermal expansion coefficients and two designs are proposed that minimise the interface stresses. These are: a 'brush'-like structure with rectangular fibres set in a Cu substrate using the 'active metal casting' (AMC) technique; and thin monoblocks (or lamellae) brazed or active metal cast onto a Cu tube. Analyses of the lamellae concept for steady-state heat loads of 5 MW m − 2 are presented. Fatigue analyses show that both solutions are theoretically viable ( 10 4 cycles). A 'brush' mock-up has been manufactured and progress on its testing is reported. Results of all tests and their relevance to the ITER design are discussed.
Fusion Engineering and Design, 2001
In a next step D/T fusion device like ITER, an intense neutron flux will be produced as a consequence of the nuclear fusion reactions. The effects of the neutron induced damage in the microstructure of the plasma-facing material (PFM) may significantly change the thermal properties and the mechanical properties as well as the behaviour of the swelling and the tritium retention in such materials. In addition, a peak heat flux as high as 20 MW m − 2 and a plasma flux of 10 18 -10 20 cm − 2 s − 1 are expected in the divertor zone during the normal operation of the reactor. The divertor materials have to withstand the neutron damage, the high heat fluxes and the high erosion caused by the interaction with the high flux plasma. The sputtered particles are co-deposited with plasma, which may contribute significantly to the total tritium inventory in the PFM. Furthermore, the interaction of steam with the sputtered particles (with usually high specific surfaces) could produce large amounts of hydrogen. All of the above topics represent critical issues for plasma performance, safety and economy, as they could limit the use of some PFM materials in next generation fusion devices. Therefore, substantial R&D effort is needed to elucidate the effects of the neutron induced damage on microstructure, erosion/deposition, tritium retention and dust formation, as well as on hydrogen production. In the framework of the European Fusion R&D program, an extensive effort on neutron effects of the material properties: namely, thermal conductivity, mechanical properties, dimensional stability, tritium trapping, erosion/deposition, co-deposition, dust formation/removal, chemical reactivity with steam and oxygen, outgassing, baking and tritium removal from PFM have been undertaken during the past several years. In this paper, : S 0 9 2 0 -3 7 9 6 ( 0 1 ) 0 0 2 5 5 -1 the recent progress achieved within the European Fusion R&D program and contributions to the development of ITER PFMs are presented and critically discussed.
Design, manufacture and initial operation of the beryllium components of the JET ITER-like wall
Fusion Engineering and Design, 2013
The aim of the JET ITER-like Wall Project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred. Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets.
Influence of Hydrogen Plasma on the Surface Structure of Beryllium
Materials
This paper presents the research results of hydrogen plasma effect on the surface structure of the TGP-56 beryllium. In the linear simulator, the operating conditions of the first wall of ITER are simulated. Beryllium was irradiated with hydrogen plasma at surface temperatures of ~360 °C, ~800 °C, and ~1200 °C, depending on its location in the ITER chamber; with a different number of pulses with a duration of each pulse of 500 s. Samples of irradiated beryllium were subjected to a set of material studies. Experimental data were obtained on the change in the structure of the surface and edges of the beryllium samples after the plasma effect. It was found that at normal (2 MW/m2) and increased (4.7 MW/m2) heat fluxes on the first wall of the ITER, the edges and beryllium surface have good resistance to erosion. Under critical conditions close to the melting point, beryllium strongly erodes and evaporates. It has been established that this material has a high resource resistance to hyd...
Nuclear Materials and Energy
In this study, beryllium tiles from Joint European Torus (JET) vacuum vessel wall were analysed and compared regarding their position in the vacuum vessel and differences in the exploitation conditions during two campaigns of ITER-Like-Wall (ILW) in 2011-2012 (ILW1) and 2013-2014 (ILW2) Tritium content in beryllium samples were assessed. Two methods were used to measure tritium content in the samples-dissolution under controlled conditions and tritium thermal desorption. Prior to desorption and dissolution experiments, scanning electron microscopy and energy dispersive x-ray spectroscopy were used to study structure and chemical composition of plasma-facing-surfaces of the beryllium samples. Experimental results revealed that tritium content in the samples is in range of 2•10 11-2•10 13 tritium atoms per square centimetre of the surface area with its highest content in the samples from the outer wall of the vacuum vessel (up to 1.9•10 13 atoms/cm 2 in ILW1 campaign and 2.4•10 13 atoms/cm 2 in ILW2). The lowest content of tritium was found in the upper part of the vacuum vessel (2.0•10 12 atoms/cm 2 and 2.0•10 11 atoms/cm 2 in ILW1 and ILW2, respectively). Results obtained from scanning electron microscopy has shown that surface morphology is different within single tile, however if to compare two campaigns main tendencies remains similar.
Selection of plasma facing materials for ITER
Proceedings of 16th International Symposium on Fusion Engineering, 1995
ITER will be the first tokamak having long pulse operation using deuterium-tritium fuel. The problem of designing heat removal structures for steady state in a neutron environment is a major technical goal for the ITER Engineering Design Activity (EDA). The steady state heat flux specified for divertor components is 5 MW/m 2 for normal operation with transients to 15 MW/m 2 for up to 10 s. The selection of materials for plasma facing components is one of the major research activities. Three materials are being considered for the divertor; carbon fiber composites, beryllium, and tungsten. This paper discusses the relative advantages and disadvantages of these materials. The final selection of plasma facing materials for the ITER divertor will not be made until the end of the EDA.
Beryllium plasma-facing components for the ITER-like wall project at JET
Journal of Physics: Conference Series, 2008
ITER-Like Wall Project has been launched at the JET tokamak in order to study a tokamak operation with beryllium components on the main chamber wall and tungsten in the divertor. To perform this first comprehensive test of both materials in a thermonuclear fusion environment, a broad program has been undertaken to develop plasma-facing components and assess their performance under high power loads. The paper provides a concise report on scientific and technical issues in the development of a beryllium first wall at JET.