Results 1 - 10 of 163
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[en] Coated conductors could be promising materials for the fabrication of the large magnet systems of future fusion devices. Two prototype conductors (flat cables in steel conduits), each about 2 m long, were manufactured using coated conductor tapes (4 mm wide) from Super Power and SuperOx, with a total tape length of 1.6 km. Each flat cable is assembled from 20 strands, each strand consisting of a stack of 16 tapes surrounded by two half circular copper profiles, twisted and soldered. The tapes were measured at 12 T and 4.2 K and the results of the measurements were used for the assessment of the conductor electromagnetic properties at low temperature and high field. The two conductors were assembled together in a sample that was tested in the European Dipole (EDIPO) facility. The current sharing temperatures of the two conductors were measured at background fields from 8 T up to 12 T and for currents from 30 kA up to 70 kA: the measured values are within a few percent of the values expected from the measurements on tapes (short samples). After electromagnetic cycling, T _c_s at 12 T and 50 kA decreased from about 12 K to 11 K (about 10%), corresponding to less than 3% of I _c. (paper)
[en] Tungsten (W) is considered as a promising candidate for plasma-facing materials for future nuclear fusion devices, and selecting optimal alloying constituents is a critical issue to improve radiation resistance of the W alloys as well as to improve their mechanical properties. We conducted in the current study a series of first-principles calculations for investigating solvent-solute mixed dumbbells in W crystals. The results suggested that titanium (Ti), vanadium (V), and chromium (Cr) are favorable as solutes for W alloys from irradiation-effect perspectives because these elements are expected to promote vacancy-interstitial recombination without causing radiation-induced precipitation that reduces ductility of irradiated materials.
[en] Reduced activation ferritic/martensitic (RAFM) steel is being developed indigenously for the Indian Test Blanket Module (TBM) structure which will be tested in International Thermonuclear Experimental Reactor (ITER). It is a modification of Grade 91 steel in which Mo and Nb are replaced by W and Ta respectively. In the present investigation fracture toughness at 0.2 mm crack extension (J0.2) and tensile properties of RAFM steel have been studied at 298, 653 and 823 K. The values of ∼ J0.2 at 298, 653 and 823 K were 106, 131 and 160 kJ.m-2 respectively, indicating no temperature dependence in this range. (author)
[en] Advanced fusion reactors structural materials (like in case of TBM and, first wall components) have several operation challenges due to the demanding high temperature exposure conditions (∼800°C) and low neutron radiation effects. The present paper reports the preliminary case studies carried out on steel and copper EMP joints and their properties characterization towards establishing this technology for ODS alloys. The EMP joints in form of tubes are fabricated and tested (typical process parameters ∼ Voltage 25 kV, Current ∼600-800 kA, Max. energy ∼ 50 kJ, and 50 sec duty cycle as major process parameters). The weld joints are further characterized by X-ray radiography and found that there were no measureable defects/discontinuities across the weld interface. This indicates the good process of joining and acceptable. Characterization studies like microstructure, interface grain orientation features, deformation, hardness has been carried out. SEM studies also carried to check the interface status and some interesting features of discontinuities are observed which are not exclusively revealed by radiography tests. Hardness survey also revealed that there is no much variation in the both parent materials as well at weld zone indicating the no hardening affects like in arc/beam weld process. EMP joining has potential features for the joining requirements of ODS kind typical metallurgical requirements
[en] WCrY Smart Alloys are developed as first wall material of future fusion devices such as DEMO. They aim at behaving like pure W during plasma operation due to depletion of the alloying elements Cr and Y. The Cr concentration gradients induced by preferential plasma sputtering cause Cr-diffusion. The exposure of WCrY and W samples to pure D plasma, with a plasma ion energy of , is simulated using the dynamic version of SDTrimSP. Cr-diffusion is included into the model. Simulation results are compared with experimental results. At sample temperatures of more than 600∘C and sputtering by D plus residual oxygen in the plasma ion flux, the Cr-transport to the surface leads to enhanced erosion for WCrY samples. A diffusion coefficient for Cr in WCrY of the order of is determined. The suitability of WCrY as first wall armour and the influence of further effects, considering especially Cr-diffusion, is discussed.
[en] The concept of the vitreous state is defined and the essential structural characteristics of glasses are discussed. A general description is given of the main properties of glasses. The various non-classical methods of obtaining glasses are presented and their modern applications indicated, particularly in the field of nuclear technology
[fr]Apres avoir defini le concept du verre on passe en revue ses caracteristiques structurales essentielles. Les proprietes des materiaux vitreux sont ensuite presentees d'une maniere synthetique. On indique les divers modes non classiques d'obtention des verres et discute les applications modernes, notamment dans le domaine nucleaire
[en] High temperature structural materials are in great demand for power, chemical and nuclear industries which can perform beyond 1000 °C as super alloys usually fail. In this regard, Mo based TZM alloy is capable of retaining strength up to 1500 °C with excellent corrosion compatibility against molten alkali metals. Hence, currently this alloy is considered an important candidate material for high temperature compact nuclear and fusion reactors. Due to reactive nature of Mo and having high melting point, manufacturing this alloy by conventional process is unsuitable. Powder metallurgy technique has limited success due to restriction in quantity and purity. This paper deals with fabrication of TZM alloy by nonconsumable tungsten arc melting technique. Initially a ternary master alloy of Mo-Ti-Zr was prepared which subsequently by dilution method, was converted into TZM alloy gradually by external addition of Mo and C in various proportions. A number of melting trials were conducted to optimize the process parameters like current, voltage and time to achieve desired alloy composition. The alloy was characterized with respect to composition, elemental distribution profile, microstructure, hardness profile and phase analysis. Well consolidated alloy button was obtained having desired composition, negligible material loss and having microstructure as comparable to standard TZM alloy. (author)
[en] The use of tungsten (W) as material for plasma-facing components (PFM) in fusion devices is reviewed with respect to its plasma and material compatibility under burning plasmas conditions. Fusion-relevant plasma operation with W walls is characterised by the need to operate at high edge densities, no or moderate density peaking, and external tools to control the W transport in the plasma core. Several surface and material issues related with the high particle fluencies in fusion devices needs further R and D but are not considered from present view to seriously limit the use of W as PFM. Reliable control of Edge Localised Modes (ELMs) and disruptions is indispensable for the application of W, both to control the W transport in the edge and to avoid target melting in uncontrolled events which can seriously detoriate the operational performance of the device. For DEMO and reactors, the behaviour of W under large neutron fluencies has to be further clarified and measures must be developed to mitigate degradation of material properties by neutron damage.