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Muta H.,Osaka University | Etoh Y.,Nippon Nuclear Fuel Development | Ohishi Y.,Osaka University | Kurosaki K.,Osaka University | Yamanaka S.,Osaka University
Journal of Nuclear Science and Technology | Year: 2012

Hydrogen diffusion in monoclinic and tetragonal zirconium oxides has been studied by electronic state calculations. In both structures, the optimized hydrogen site lies near the center of a distorted fluorite structure. The activation energy was calculated to be 120-200 kJ/mol, which is similar to experimentally measured values. The effects of compressive stress, alloying elements, and oxygen defects are considered individually. Compressive stress reduces the hydrogen diffusion coefficient by 40%/GPa. Oxygen defects and substituted Fe and Cr are thought to act as trapping sites for hydrogen, which probably reduces hydrogen diffusion in zirconium oxide. © 2012 Atomic Energy Society of Japan.

Ishiyama Y.,Nippon Nuclear Fuel Development | Rogge R.B.,National Research Council Canada | Obata M.,Toshiba Corporation
Journal of Nuclear Materials | Year: 2011

Weld beads on plate specimens made of type 316L stainless steel were neutron-irradiated up to about 2.5 × 1025 n/m2 (E > 1 MeV) at 561 K in the Japan Material Testing Reactor (JMTR). Residual stresses of the specimens were measured by the neutron diffraction method, and the radiation-induced stress relaxation was evaluated. The values of σx residual stress (transverse to the weld bead) and σy residual stress (longitudinal to the weld bead) decreased with increasing neutron dose. The tendency of the stress relaxation was almost the same as previously published data, which were obtained for type 304 stainless steel. From this result, it was considered that there was no steel type dependence on radiation-induced stress relaxation. The neutron irradiation dose dependence of the stress relaxation was examined using an equation derived from the irradiation creep equation. The coefficient of the stress relaxation equation was obtained, and the value was 1.4 (×10-6/MPa/dpa). This value was smaller than that of nickel alloy. © 2010 Elsevier B.V. All rights reserved.

Fukuya K.,Japan Institute of Nuclear Safety System | Nishioka H.,Japan Institute of Nuclear Safety System | Fujii K.,Japan Institute of Nuclear Safety System | Miura T.,Japan Institute of Nuclear Safety System | Torimaru T.,Nippon Nuclear Fuel Development
Journal of Nuclear Materials | Year: 2011

Local deformation behavior in cold worked SUS316 stainless steels irradiated to 73 dpa at 573 K was examined by SEM and EBSD after deformation to plastic strain of ∼2% at 593 K at a strain rate of 7.8 × 10 -8/s in an argon gas environment. Grain boundary separation occurred at random grain boundaries which lay almost normal to the tensile direction and had coarse dislocation channels in grains on one side. Misorientation maps indicated that a local high strain field existed near grain boundaries when coarse channels impinged on grain boundaries forming dislocation pileups. The grain boundary separation occurred more frequently at the grain boundaries where the difference in Schmidt factor was larger between two adjacent grains. These results indicated that the grain boundary separation was triggered by the local high stress field induced by dislocation pileups and high normal component of tensile stress. © 2011 Elsevier B.V. All rights reserved.

Kubo T.,Nippon Nuclear Fuel Development | Kobayashi Y.,M.O.X. Co.
Journal of Nuclear Materials | Year: 2013

Delayed hydride cracking (DHC) of Zircaloy-2 is one possible mechanism for the failure of boiling water reactor fuel rods in ramp tests at high burnup. Analyses were made for hydrogen diffusion around a crack tip to estimate the crack velocity of DHC in zirconium alloys, placing importance on effects of precipitation of δ-hydride. The stress distribution around the crack tip is significantly altered by precipitation of hydride, which was strictly analyzed using a finite element computer code. Then, stress-driven hydrogen diffusion under the altered stress distribution was analyzed by a differential method. Overlapping of external stress and hydride precipitation at a crack tip induces two stress peaks; one at a crack tip and the other at the front end of the hydride precipitate. Since the latter is larger than the former, more hydrogen diffuses to the front end of the hydride precipitate, thereby accelerating hydride growth compared with that in the absence of the hydride. These results indicated that, after hydride was formed in front of the crack tip, it grew almost steadily accompanying the interaction of hydrogen diffusion, hydride growth and the stress alteration by hydride precipitation. Finally, crack velocity was estimated from the calculated hydrogen flux into the crack tip as a function of temperature, stress intensity factor and material strength. There was qualitatively good agreement between calculation results and experimental data. © 2013 Published by Elsevier B.V.

Fukuya K.,Japan Institute of Nuclear Safety System | Nishioka H.,Japan Institute of Nuclear Safety System | Fujii K.,Japan Institute of Nuclear Safety System | Miura T.,Japan Institute of Nuclear Safety System | Kitsunai Y.,Nippon Nuclear Fuel Development
Journal of Nuclear Materials | Year: 2013

Local strain near grain boundaries under tensile stress was examined using the EBSD technique for cold-worked SUS316 stainless steel irradiated to 73 dpa. Distribution of misorientation in the same areas was analyzed while macroscopically deforming the specimen at an elastic strain level of 0.03% and a plastic strain level of 3%. A clear increase in local strain was detected within 4 μm from the grain boundaries at the 3% plastic strain. It was confirmed that high local strain was produced at the 0.03% elastic strain near the grain boundaries which exhibited higher misorientations at the plastic strain. Detailed analysis in areas within 1 μm from individual grain boundaries also revealed that local strain was increased at some boundaries at the 0.03% elastic strain. © 2012 Elsevier B.V. All rights reserved.

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