Showing papers in "Journal of Nuclear Materials in 2013"

TL;DR: In this article, the authors consider the risks engendered by the baseline divertor strategy with regard to known W plasma-material interaction issues and briefly present the current status of a possible full-tungsten (W) divertor design.

TL;DR: In this article, the progress of work within the EFDA long-term fusion materials program in the area of tungsten alloys is reviewed, with a detailed overview of the latest results on materials research, fabrication processes, joining options, high heat flux testing, plasticity studies, modelling, and validation experiments.

TL;DR: In this paper, a comparison of a range of commercial and model alloys, conventional austenitic steels do not have sufficient oxidation resistance with only ∼18Cr-10Ni, and higher alloyed type 310 stainless steel is protective but Ni is not a desirable alloy addition for this application.

TL;DR: In this paper, the fracture behavior is improved by using tungsten laminated materials and wire reinforced materials, which can achieve self-passivation, which is essential in case of loss-of-coolant accidents for plasma facing materials.

TL;DR: In this paper, a new interatomic pair potential for W-He is described, which includes a short range modification to the Ackland-Thetford tungsten potential, and molecular dynamics simulations using these potentials accurately reproduce ab initio results of the formation energies and ground state positions of He point defects and self interstitial atoms.

TL;DR: In this paper, a survey of properties data for Zirconium carbide (ZrC) is provided in support of the current efforts toward fuel performance modeling and providing guidance for future research on ZrC for fuel applications.

TL;DR: In this article, the authors present recent RAFM steel results obtained in ITER partner countries in relation to different TBM and DEMO options, and evaluate the mechanical properties of these steels before and after irradiation and in contact with different cooling media.

TL;DR: In this paper, the ITER-like wall (ILW) was installed in the JET to enable a direct comparison of operation with all carbon plasma facing components (PFCs) to an all metal beryllium/tungsten first-wall under identical conditions.

TL;DR: In this paper, the authors describe a methodology to examine dislocation loops in irradiated steels based on a combination of crystallographic information and g⋅b invisibility criteria.

TL;DR: In this article, the authors used nanoindentation to measure the change in hardness as a function of six damage levels (0, dpa, 0.07, 0.4, 1.2, 13,dpa, and 33 dpa).

TL;DR: In this article, the authors investigated the variation in displacements per atom (dpa) and helium production levels as a function of position within the high flux regions of a recent conceptual model for the next-step fusion device DEMO.

TL;DR: In this article, a physically-based, empirically calibrated model for estimating irradiation-induced transition temperature shifts in reactor pressure vessel steels, based on a broader database and more complete understanding of embrittlement mechanisms than was available for earlier models, is presented.

TL;DR: Tungsten erosion in the outer divertor of the JET ITER like wall was quantified by spectroscopy as mentioned in this paper, and the signature of prompt redeposition was observed in the analysis of WI 400.

TL;DR: In this article, the viability of advanced oxidation-resistant Fe-base alloys to protect zirconium from rapid oxidation in high-temperature steam environments has been examined, and the applicability of these protective layers in light-water-reactor nuclear fuel structures is offered.

TL;DR: In this paper, the authors investigated the corrosion characteristics on several selected alloys at 600 and 700°C in FLiNaK molten salts with different moisture contents, and the results of structural characterization revealed that the tested alloys with higher moisture content would aggravate intergranular corrosion and pitting.

TL;DR: In this paper, the effect of a hypothetical reduction of the SOL power width on the performance of the ITER divertor was analyzed using the SOLPS4.3 code. And the authors found that a reduction of this width by a factor 3 (down to 1.2

TL;DR: In this paper, a metal matrix composite with good mechanical property and thermal neutron absorbing ability was investigated based on B4C/Al neutron radiation shielding material, which is used for the criticality safety during the storage or transportation of spent nuclear fuel.

TL;DR: The microstructure of four neutron irradiated Fe-Cr model alloys of industrial purity (Fe 2.5%Cr, Fe 5%Cr and Fe 9%Cr) has been characterized by atom probe tomography (APT) as mentioned in this paper.

TL;DR: In this article, the evolution of the oxide population in 14YWT and 9CrODS steels was analyzed using both transmission electron microscopy and atom probe tomography (OPT).

TL;DR: A 14YWT nanostructured ferritic alloy (NFA) was implanted with He + ions to fluences of 6.75 × 1021 He m-2 at 400 °C in order to simulate the effects of high He concentrations produced in advanced...

TL;DR: In this article, a detailed residual stress and phase fraction analysis was carried out for the oxides formed on Zircaloy-4 after autoclave exposure at 360°C for various times by means of synchrotron X-ray diffraction.

TL;DR: In this paper, the Spark Plasma Sintering (SPS) was used to sintered UO2-SiC composites at a range of temperatures from 1400 to 1600°C.

TL;DR: Operation with all tungsten plasma facing components has become routine in ASDEX Up- grade and the long term fuel retention was reduced by more than a factor of 5 as demonstrated in gas balance as well as in post mortem analyses.

TL;DR: In this paper, an overview of the edge-localized mode (ELM) control techniques currently being developed is discussed along with the requirements for applying these techniques to plasmas in ITER.

TL;DR: In this paper, a modified melt-based process was used to fabricate breeder pebbles with additions of titania in order to obtain lithium metatitanate as a secondary phase.

TL;DR: In this article, the hardness of pure Fe and model Fe-Cr alloys containing 5, 10 and 14%Cr were irradiated with Fe+ ions at a maximum energy of 2.5 MeV to the same dose of 0.6 MeV at temperatures of 300, 400, and 500°C.

TL;DR: In this article, the effects of tungsten and tantalum contents on impact, tensile, low cycle fatigue and creep properties of reduced activation Ferritic-Martensitic (RAFM) steel were studied to develop India-specific RAFM steel.

TL;DR: The Ringhals Units 3 and 4 PWRs in Sweden are pressurized water reactors (PWRs) designed and supplied by Westinghouse Electric Company, with commercial operation in 1981 and 1983, respectively as discussed by the authors.

TL;DR: In this article, the effect of radiation damage by heavy ions at cryogenic and elevated temperatures on zirconium nitride was evaluated. But, the effect on the defect migration was not observed at low temperatures nor bubble formation observed at elevated temperatures.

TL;DR: Corrosion tests were performed in steam and supercritical water at 500 °C for two ferritic-martensitic alloys: HCM12A and NF616 as mentioned in this paper.