Materials Science Forum Vols. 475-479

Abstract: In the development of nuclear fusion into a future peaceful energy source, the scheme of magnetically confined fusion plasmas has experienced substantial progress in the last decades. Some of the main remaining materials issues are the high heat loads and the erosion of plasmafacing components by erosion from particle impact. Since component lifetime considerations and the negative influence of impurities on the plasma lead to opposite conclusions about the materials choice, an experimental investigation of the applicability of tungsten as plasma-facing material is currently performed at the tokamak fusion experiment ASDEX Upgrade in Garching, Germany. For this purpose an industrial scale method to coat graphite tiles with tungsten thin films was required. Therefore a comparative study of the heat load behaviour of coatings deposited with various PVD techniques was performed. Tungsten coatings on graphites and fiber reinforced carbons were investigated with respect to their properties relevant for fusion research including electron-beam loading with heat loads exceeding 20MW/m2.

Abstract: This paper reported the low activation martensitic steels which are being studied to develop the structural materials in fusion reactors. The steels were based on 9Cr1.5WVTa, but the effect of alloy elements was investigated by changing the amounts of alloy elements or adding other elements. The structure and properties of the steels were studied by tensile experiment, X-ray diffraction, SEM, TEM. Also the metallurgical process and heat treatment effect were discussed.

Abstract: Intergranular attack/stress corrosion cracking of Alloy 600 continues to be an issue in the tube/tube support plate crevices and top of tubesheet locations of recirculating steam generators and in the upper bundle of free span superheated regions of once through steam generators (OTSG). Recent examinations of degraded pulled tubes from several plants suggest possible lead involvement in the degradation. Laboratory investigations have been performed to determine the factors influencing lead cracking in Alloy 600 and Alloy 690 steam generator tubes. The test environment is believed to be prototypical, with the addition of lead oxide, of a concentrated liquid phase existing in the pores of thin deposits on upper bundle tubes of an OTSG. Highly strained reverse U-bend specimens were tested at controlled electrochemical potentials. Maximum susceptibility was at open circuit potential, unlike cracking of Alloy 600 in caustic and acid sulfate environments where maximum susceptibility occurs when specimens are polarized above the open circuit potential. Transgranular, intergranular and mixed mode cracking was observed and in all Alloy 600 conditions tested (mill annealed, sensitized, thermally treated) while thermally treated Alloy 690 has so far resisted cracking. A film rupture/anodic dissolution model with displacement plating of Pb preceding passive film formation is consistent with the experimental observations

Abstract: We report the formation of anomalous defect clusters in yttria stabilized zirconia (YSZ: 13 mol% Y2O3-ZrO2), which are a primary candidate for inert matrix fuels. Electron irradiation with 100 to 1000 keV is found to induce characteristic defect clusters. The defect clusters show strong black/black lobes contrast and multiply dislocations when the defect clusters grow to a critical diameter of 1.0-1.5 µm. Our analysis with transmission electron microscopy and numerical calculations lead to a conclusion that these defect clusters are charged oxygen platelets formed by the selective displacements of oxygen sublattice.

Abstract: Research on low cyclic fatigue behaviors of two kinds of Zr-alloy(1#:Zr-1Sn-0.3Nb- 0.3Fe-0.1Cr.2#: Zr-1Sn-1Nb-0.4Fe-)at 400°C,The fractural characterization was investigated by SEM.

Abstract: The effects of annealing at 570oC and 640oC on the microstructural and corrosion characteristics for Zr-1.0Nb-1.0Sn-0.1Fe alloy were elucidated. After annealing at 570oC below the temperature of a monotectoid reaction in the Zr-Nb system, both orthorhombic Zr3Fe and the bcc b-Nb particles were uniformly found and the mean size of the second phase particles was increased with an increasing of the annealing time. In the case of an annealing at 640oC for 2 h above the monotectoid reaction temperature, the Zr3Fe was observed intermittently and after a longer annealing of 1000 h the b-Zr particles were well developed. The corrosion resistance after the 570oC anneal was improved as the annealing time increased, while that after the 640oC anneal decreased as the annealing time increased. The fraction of the tetragonal phase within the ZrO2 oxide increased as the corrosion resistance was improved. It was concluded that the equilibrium Nb concentration and the formation of the tetragonal ZrO2 due to the b-Nb phase would lead to improving the corrosion resistance of the alloy.

Abstract: The creep resistance of one kind of Zirconium based alloy with Niobium, Iron, Chromium , Tin alloying elements was investigated, the creep curve ( e-t) was obtained in stress levels of 117MPa 、137MPa、157MPa respectively by holding 200 hours at 400°C. The results show that the creep resistance of this alloy is much better than that of Zircaloy-4. The microstructure of post- creep test was observed by TEM, which shows that the dislocation morphology is changed from dislocation lines to dislocation cells with strain increasing , and some precipitates grow larger. It is concluded that the creep is controlled by dislocation climb-glide diffusion mechanism.

Abstract: The aim of this study is to investigate a change in delayed hydride cracking (DHC) velocity of Zr-2.5Nb tubes with fast neutron fluence (E>1MeV) and predict the DHC velocity of the irradiated Wolsong 1 Zr-2.5Nb tubes at a neutron fluence corresponding to the 30 year design lifetime. To this end, the DHC velocity were determined at temperatures ranging from 100 to 280 oC on unirradiated Zr-2.5Nb tubes and the irradiated Zr-2.5Nb tubes in the Wolsong Unit-1 to the neutron fluence of 8.9x1025 n/m2 (E>1MeV). DHC tests were conducted on the compact tension specimens charged with 34 to 100 ppm hydrogen in accordance with the KAERI DHC procedures that have been validated through a round robin test on DHC velocity of Zr-2.5Nb tubes as an IAEA coordinated research project. Irradiated Zr-2.5Nb tubes had 3 to 5 times higher DHC velocity than that of unirradiated Zr-2.5Nb tubes while the inlet region of the irradiated Zr-2.5Nb tube with the highest yield strength had a slightly higher DHC velocity compared to that of the outlet region with the lowest yield strength. From a normalized correlation of yield strength and DHC velocity of the Zr-2.5Nb tubes, the yield strength was found to govern the DHC velocity of the Zr-2.5Nb tubes irrespective of the neutron fluence and operating temperatures. The DHC velocity of the irradiated Zr-2.5Nb tubes is predicted after a 30 year operation in the Wolsong Unit 1 on the basis of an increase in the yield strength with neutron fluence and a DHC velocity dependence on the yield strength of Zr-2.5Nb tubes.

Abstract: The small punch (SP) test is a novel technique that uses a relatively small volume of material, to enable a measure of load versus deflection to determine the mechanical behaviour of the alloy. This study aims to investigate the mechanical properties of Zircaloy-4 (Zr-4) alloy which will be used for construction of many of the core components in ANSTO’s replacement research reactor at Lucas Heights in Australia. The Zr-4 alloy was hydrided under different conditions to simulate the radiation-induced reduction in fracture toughness over the service life of the reactor. The SP test was applied to determine the deformation and fracture behaviour of the hydrided materials. It was found that as the hydride formation increased, the equivalent fracture strain (eqf) of the alloy decreased to lower values.

Abstract: Both small experimental extruded tubes and full-size pressure tubes were examined using scanning electron microscope/electron backscattered diffraction (SEM/EBSD) and transmission electron microscope/selected area diffraction (TEM/SAD). The final microstructures and textures vary with billet microstructure, extrusion temperature and extrusion ratio. Three components in {0002} pole figures were determined. The first component (radial) is produced by a and c+a slip in a-grains during extrusion. The second component (transverse) is associated with the elongated a- grains with their c-axes parallel to their long dimension. The third component (axial) is produced by b-a phase transformation after extrusion with a preferred variant of the Burgers relationship.