Showing papers in "Journal of Nuclear Materials in 2007"

TL;DR: In this paper, a compilation of non-irradiated and irradiated properties of SiC are provided and reviewed and analyzed in terms of application to TRISO fuels, specifically in the high-temperature irradiation regime.

TL;DR: In this article, the authors investigated the use of elevated-temperature ferritic/martensitic steels for in-core and out-of-core applications for the next generation of nuclear power reactors.

TL;DR: In this article, the current understanding of the response of candidate materials for SCWR systems, focusing on the corrosion and stress corrosion cracking response, and highlights the design trade-offs associated with certain alloy systems.

TL;DR: In this paper, the authors review the development, characterization, and irradiation effect studies for SiC composites for fusion energy applications and make the case that SiC composite are progressing from the stage of potential viability and proof-of-principle to one where they are ready for system demonstration, i.e., for flow channel inserts in Pb-Li blankets.

TL;DR: A series of R&D activities on the structural material China Low Activation Martensitic steel (CLAM) and related blanket technology are being carried out in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as mentioned in this paper.

TL;DR: Atom probe tomography has played a key role in the understanding of the embrittlement of neutron irradiated reactor pressure vessel steels through the atomic level characterization of the microstructure as mentioned in this paper.

TL;DR: The US experience with mixed oxide, metal, and mixed carbide fuels is substantial, comprised of irradiation of over 50,000 MOX rods, over 130,000 metal rods, and 600 mixed-carbide rods, in EBR-II and FFTF alone as mentioned in this paper.

TL;DR: In this article, four ferritic alloys based on the composition Fe-14Cr-3W-0.4Ti were developed with a predominant dispersion of either oxide particles or nano-size Y-, Ti-, O-rich clusters, or nanoclusters (NC).

TL;DR: In this paper, the authors reviewed recent research results obtained in Europe, Japan, China and USA on reduced activation ferritic/martensitic (RAFM) steels and presented a strong case for the use of the steels in ITER test blanket modules.

TL;DR: A series of liquid lithium lead blankets have been designed and assessed in China, with four types of liquid lead blankets including the RAFM steel-structured He-cooled quasi-static LiPb tritium breeder (SLL) blanket.

TL;DR: In this paper, the authors assess the range of applicability of the available interatomic potentials for UO 2 by static calculations, together with defect properties: the formation and activation energies of vacancies and interstitials, the binding energy of small clusters of these defects and the volume change associated with them.

TL;DR: In this article, the effects of neutron irradiation on microstructural evolution and radiation hardening were examined for fine-grained W-0.3%TiC and commercially available pure W-20μm.

TL;DR: In this article, three different tungsten alloys (pure W, a lanthanum-oxide dispersion strengthened W-alloy and a potassium doped Tungsten alloy) in different fabrication conditions (sintered and rolled rods) were tested to characterize their fracture behaviour at different temperatures (from −196 to 800 °C).

TL;DR: The main directions of the project oriented materials activity and main challenges related to selection of materials for the ITER components are discussed in this article, where the justification of materials performance is fully documented, which allows traceability and reliability of design data.

TL;DR: In this article, the oxidation behavior of ferritic-martensitic (F-M) alloys in supercritical water (SCW) was studied in order to evaluate the suitability of these alloys for use in the Gen IV super-critical water reactor (SCWR) concept.

TL;DR: The GLADIS ion beam facility at IPP Garching is equipped with two individual 1.1 MW power ion sources for testing actively cooled plasma facing components under high heat fluxes as discussed by the authors.

TL;DR: In this article, a he-cooled divertor concept has been investigated at Forschungszentrum Karlsruhe in cooperation with the Efremov Institute.

TL;DR: In this article, an investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER is concerned.

TL;DR: The effects of fast neutron irradiation on SiC and SiC composites have been studied in this paper, where a statistically significant population of flexural samples were irradiated up to 4.6 × 1025 n/m2 (E < 0.1 ) at 300, 500, and 800 °C in the High Flux Isotope Reactor at Oak Ridge National Laboratory.

TL;DR: In this article, the corrosion behavior of model Zr-based alloys at 500°C was assessed by long term (up to 400 days) corrosion testing in an effort to evaluate their potential for use in the supercritical water reactor and to assess the influence of alloying elements on corrosion behavior.

TL;DR: In this paper, a thermo-mechanical treatment (TMT) was developed that increased the strength from room temperature to 700°C of commercial nitrogen-containing steels and new steels designed for the TMT.

TL;DR: In this article, the authors measured the trapping of hydrogen and helium in polycrystalline tungsten irradiated with 500 eV He +, H + and D + ions individually or sequentially.

TL;DR: The history of Molten salt Reactor technology reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed as mentioned in this paper.

TL;DR: An overview of the new ITER-like wall project in JET is presented in this article, which aims at an optimal use of JET's unique features: physical size, plasma parameters most closely to ITER and the capability to handle beryllium and tritium, allowing the study of critical questions related to operating within the limits of ITER wall materials.

TL;DR: In this paper, the most important questions that must be addressed to demonstrate the feasibility of a Molten Salt Reactor (MSR) have been reviewed, and they have been shown that the MSR can be used as either waste incinerators or thorium cycle systems.

TL;DR: In this paper, a critical review of old and new data reported in the open literature on the main thermo-physical properties of the molten Pb and Pb-Bi: characteristic temperatures, latent heats of melting and evaporation, surface tension, density, heat capacity, viscosity, electric and thermal conductivity.

TL;DR: In this paper, the nano-oxides of the K1 were cubic pyrochlore Y2Ti2O7 while those of K4 were mainly perovskite AlYO3, which was confirmed by transmission electron microscopy and X-ray diffraction (XRD) analyses.

TL;DR: In this paper, the effects of ion-irradiation on the swelling behavior were investigated for a high-chromium oxide dispersion strengthened (ODS) ferritic steel and a 9Cr-2W reduced-activation ferritic (RAF) steel up to a high dose (nominal 60 dpa) with and without helium (900 appm) by using single and dual-ion irradiation techniques at 773 K.

TL;DR: In this paper, the poloidal variation of the parallel flow has been measured by moving the contact point of a small circular plasma onto limiters at different poloidal angles, and the resulting variations of flow are consistent with the existence of a poloidally localized enhancement of radial transport concentrated in a 30° sector near the outboard midplane.

TL;DR: In this paper, a new hydrogen isotope barrier material, erbium oxide Er2O3, or erbia, has been tested with thermal load test and permeation measurements to determine whether erbia is compatible with other materials at elevated temperatures and if it suppresses hydrogen transport.