Showing papers in "Fusion Engineering and Design in 2011"

TL;DR: A review of the development of edge plasma modeling at ITER and of its interaction with the evolving divertor design is presented in this article, where the SOLPS (B2-Eirene) code has been developed for, and applied to, the evaluation and the design of the ITER divertor for the last 15 years.

TL;DR: In this article, the authors investigated the irradiation hardening of Fe-based model ferritic alloys after Fe-ion irradiation experiments in order to deduce mechanistically based nominal hardness from the nano-indentation tests on the ion-irradiated surface.

TL;DR: In this article, a detailed pass-by-pass finite element simulation is presented to investigate the residual stresses in narrow gap multipass welding of pipes with a wall thickness of 70mm and 73 weld passes.

TL;DR: In this paper, the effect of heat input on microstructure of indigenously developed reduced activation ferritic martensitic (RAFM) steel weld metal produced by GTAW process has been studied.

TL;DR: In this paper, the authors used diffusion boning to join tungsten and EUROFER97 for the manufacturing of some components in the fusion technology, where a Nb plate is introduced as an interlayer to reduce the interfacial residual stresses induced during diffusion bonding by the large mismatch of their coefficients of thermal expansion.

TL;DR: In this article, a functionally graded tungsten/steel joint is analyzed using the finite element method and considering a simplified model of the proper divertor component, and the resulting stresses and deformations are then evaluated to determine the optimal joining parameters.

TL;DR: Recent US efforts in this area, including optimization and assessment of the different concepts under power plant conditions, are summarized to help guide future R&D, including fabrication, joining, material behavior under the fusion environment and impact of design choice on reliability.

TL;DR: The solid breeder blanket concept proposed by the European Union features the tritium breeding ceramics as pebble beds inside several breeder units as mentioned in this paper, but the pebbles need to withstand forces originating from thermal expansion mismatches and neutron irradiation.

TL;DR: In this article, a global variance reduction technique called MAGIC was proposed for determining the a priori flux information in the ITER reference neutronics model, achieving an efficiency of 74 times that of an analogue run and 48 times better than analogue.

TL;DR: In this paper, the Tore Supra platform was equipped with a full tungsten divertor, benefitting from the unique long pulse capabilities, the high installed RF power and the long experience with actively cooled high heat flux components.

TL;DR: In this article, the authors reviewed the BA R&D status of reduced activation ferritic/martensitic (RAFM) steel development in Japan, especially F82H (Fe 8Cr 2W-V, Ta), and introduced the key technical issues relevant to the design and fabrication of the DEMO blanket and the recent achievements in Japan.

TL;DR: In this paper, a short overview on tungsten (W) coatings deposited by various methods on carbon materials (carbon fibre composite and fine grain graphite) is given, and a particular attention is paid to the Combined Magnetron Sputtering and Ion Implantation (CMSII) technique.

TL;DR: In this paper, laser-based techniques in combination with spectroscopy are proposed for in situ characterization of the first wall in ITER, which can be used to determine the stored amount of tritium (tritium retention) and to characterize the layer deposition on the wall components in fusion devices.

TL;DR: In this article, the effects of protection cover and membrane thickness on the durability of Ionic-liquid-i-OMs and the efficiency of isotope separation have been evaluated, and it was observed that the durability was improved by a nafion 324 overcoat.

TL;DR: The EU Breeding Blanket Programme aims the testing of two blankets concept in ITER in form of Test Blanket Modules in the equatorial port #16 the two EU TBMs will be exposed to the plasma and the complex system of their auxiliary systems dedicated to heat and Tritium removal will be integrated in the surrounding ITER buildings as mentioned in this paper.

TL;DR: The 9Cr-ODS steel has a tempered martensitic structure with uniform oxide particles in the matrix, and the mean diameter and density of the oxide particles did not change significantly by different creep temperatures and stresses, suggesting the stability of the microstructure as mentioned in this paper.

TL;DR: In this paper, a double-layered annulus metal hydride bed for the storage and delivery of tritium is presented. But the performance of the delivery and storage is not evaluated.

TL;DR: In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R&D program as mentioned in this paper, which consisted in its initial phase in the high heat flux (HHF) testing of W mock-ups and medium scale prototypes up to 20MW/m2 in the AREVA FE 200 facility (F).

TL;DR: The COMPASS tokamak, a divertor device with clear H-mode and ITER-relevant geometry, has been reinstalled in IPP Prague after its transport from CCFE in UK.

TL;DR: In this article, the aging properties of reduced activation ferritic/martensitic steel F82H were investigated at temperature ranging from 400°C to 650°C up to 100,000h.

TL;DR: In this article, the authors present a he-cooled modular divertor with multiple-jet cooling for DEMO with the goal of reaching a heat flux of 10MW/m2.

TL;DR: The Helium Cooled Lithium Lead Test Blanket Module (HCLL-TBM) is one of the two European TBMs, representative of DEMO blanket technology, that shall be ready for installation in ITER starting from the first day of operation as discussed by the authors.

TL;DR: In this paper, the welding CCT diagram of CLAM steel was constructed and it showed that there were ferrite and martensite transformation regions only, and the diagram made it possible to predict the microstructures and properties of HAZ and became an important tool to evaluate weldability.

TL;DR: In this article, a toroidal peak of the poloidal halo current was identified as the cause of asymmetric EM loads on tokamak components, and the related loads were specified for likely, unlikely and extremely unlikely AVDEs.

TL;DR: Tungsten insert gas (TIG) welding on China low activation martensitic (CLAM) steel under identical conditions was performed as discussed by the authors, where microhardness test, tensile test, Charpy impact test and microstructure measurements were carried out on TIG welded joints after post weld heat-treatment.

TL;DR: In this paper, neutral beam heating in EAST tokamak is simulated using NUBEAM code and the NBI (Neutral Beam Inject) system of EAST has two deuterium beam lines, in which beam power is 4MW and energy is 80 keV.

TL;DR: In this paper, the authors proposed simplification of blanket structure without decreasing the net Tritium Breeding Ratio (TBR), which is necessary for a fusion DEMO reactor.

TL;DR: In this paper, the effects of frequency profiles on the sharpness of edge rounding were investigated, and it was shown that tungsten ablation with accurate electro-chemical molding is very sensitive to the kind of electric current.

TL;DR: The Karlsruhe Institute of Technology (KIT) designed, constructed and successfully tested a 68kA superconductor current lead demonstrator for ITER which led to the decision by ITER IO to use HTS current leads for the ITER magnet system, and the Chinese Domestic Agency is in charge of providing these leads as a part of the superconducting feeder system as discussed by the authors.

TL;DR: In this paper, the negative deuterium ion beam halo was investigated using a classic plasma grid, and it was shown that significant halo originates from negative ions being formed on the metal parts of the plasma grid (as could happen when neutral atoms from the source hit the plasma grids covered with caesium).