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Une K.,Nippon Nuclear Fuel Development | Takagi I.,Kyoto University | Sawada K.,Kyoto University | Miyamura S.,Kyoto University | Aomi M.,Global Nuclear Fuel Japan
Progress in Nuclear Energy | Year: 2012

In order to clarify the hydrogen diffusion mechanism in the oxide layer of zirconium alloys, in situ hydrogen isotope diffusion in the oxide layer has been examined. The zirconium alloys used were Zircaloy-2, GNF-Ziron (Zircaloy-2 type alloy with high iron content) and VB (zirconium-based alloy with high iron and chromium contents). They were corroded in 1 or 0.1 M LiOH-containing water at 563 K, producing oxide layers of 1.1-2.1 μm in thickness. The diffusion experiments were carried out in the temperature range from 488 to 633 K by using a combined technique of deuterium plasma exposure and nuclear reaction analysis for D ( 3He,p) 4He reaction. From the transient deuterium profiles in the oxide layers, it was concluded the LiOH-water-corroded oxides had a single-layer structure, which was in contrast to the double-layer structure previously observed in steam-corroded oxide layers. The diffusion coefficients in the 1 M LiOH-water-corroded oxides evaluated from the deuterium profiles were smaller in the order of Zircaloy-2 > GNF-Ziron > VB at 573 K. For the 0.1 M LiOH-water-corroded oxide of GNF-Ziron, the diffusivity was lower than that of the 1 M LiOH-water-corroded oxide by a factor of 1/4. The present diffusion coefficients of the 1 M LiOH-water-corroded oxides of GNF-Ziron and VB were approximately 7 times larger than the previous data of the corresponding steam-corroded oxides. The deuterium diffusion properties in the oxides of the three alloys obtained in the in situ experiment were roughly consistent with their hydrogen absorption performances in the LiOH-water-corrosion tests, as well as in the previous steam corrosion tests. © 2011 Elsevier Ltd. All rights reserved.


Sakamoto K.,Nippon Nuclear Fuel Development | Une K.,Nippon Nuclear Fuel Development | Aomi M.,Global Nuclear Fuel Japan | Otsuka T.,Kyushu University | Hashizume K.,Kyushu University
Journal of Nuclear Science and Technology | Year: 2015

The change of chemical states of niobium with oxide growth was examined in the oxide layers of Zr-2.5Nb around the first kinetic transition by the conversion electron yield -X-ray absorption near-edge structure measurements. The detailed depth profiles of niobium chemical states were obtained in both the pre-and the post-transition oxide layers of Zr-2.5Nb formed in water at 663 K for 40-280 d. The depth profiling revealed that the inner oxide layer remained protective to oxidizing species even though in the post-transition region and this excellent stability of barrierness would be attributed the suppression of hydrogen pickup. © 2015 Atomic Energy Society of Japan. All rights reserved.


Matsunaga J.,Nippon Nuclear Fuel Development | Matsunaga J.,Global Nuclear Fuel Japan | Matsushima K.,Nippon Nuclear Fuel Development | Hirai M.,Nippon Nuclear Fuel Development
Journal of Nuclear Science and Technology | Year: 2015

For the management of severe accidents of sodium-cooled fast breeder reactor, the coolability of the fuel debris bed on a core support plate is a key concern during the post-accident heat removal phase. In an air ingress scenario, the reactions between the fuel and highly oxidized sodium are likely to form sodium uranoplutonate. This would negatively influence the coolability of the fuel debris bed due to a lowering of the thermal conductivity and density. This study has focused on the formation kinetics of sodium uranate from UO2 and liquid sodium including oxygen at a high concentration. In this paper, the experiments on reaction initiation temperatures, reaction rates, and the decomposition of sodium uranate are reported. © 2015 Atomic Energy Society of Japan. All rights reserved.


Une K.,Nippon Nuclear Fuel Development | Takagi I.,Kyoto University | Sawada K.,Kyoto University | Watanabe H.,Kyoto University | And 2 more authors.
Journal of Nuclear Materials | Year: 2012

Deuterium diffusion in proton-irradiated oxide layer of zirconium alloy has been in situ examined at 573 K by using a combined technique of deuterium plasma exposure and nuclear reaction analysis of the D( 3He,p) 4He reaction. The zirconium alloy used was GNF-Ziron (a high iron Zircaloy-2 type alloy), which had been corroded in high temperature steam, producing an oxide layer of 1.6-1.7 μm thickness. Oxidized specimens were irradiated at ambient temperature with 350 keV H + ions, and the total fluence was 1 × 10 17 cm -2. An outer non-protective oxide layer of 0.5-0.6 μm thickness, which was observed in the unirradiated oxide layer, evolved into the protective barrier oxide due to the proton irradiation. The evaluated diffusion coefficients in the barrier layer were almost identical for both the unirradiated and irradiated oxides. From X-ray diffraction measurements, lattice expansion and high compressive stress were found in the proton-irradiated oxide. The most probable mechanism for evolution of the diffusion property in the irradiated oxide was ascribed to the increase of the compressive stress induced by the constraint of the damage-accumulated oxide layer by the thick metal substrate. © 2011 Elsevier B.V. All rights reserved.


Matsunaga J.,Nippon Nuclear Fuel Development | Une K.,Nippon Nuclear Fuel Development | Kusagaya K.,Global Nuclear Fuel Japan
LWR Fuel Performance Meeting/Top Fuel/WRFPM 2010 | Year: 2010

A chemical trap demonstration was conducted for the Aluminosilicate(Al-Si- O) Additive Fuel, and SEM and TEM observation was carried out. The Iodine reacted with Cs-doped Al-Si-O Additive Fuel and forms CsI on UO2 grain boundary. The thermodynamic calculation indicated CsI can form from cesiumsilicate even if the iodine activity is too low to form zirconiumiodide. The numerical simulation indicated iodine diffusion can be delayed by the existence of Cs-Al-Si-O phase. These results pointed to a mechanism providing higher PCI-SCC resistance by the Al-Si-O Additive Fuel.

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