Kawasaki, Japan
Kawasaki, Japan

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Nemoto Y.,Japan Atomic Energy Agency | Kaji Y.,Japan Atomic Energy Agency | Ogawa C.,Japan Atomic Energy Agency | Kondo K.,Japan Atomic Energy Agency | And 3 more authors.
Journal of Nuclear Materials | Year: 2017

The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study. © 2017 Elsevier B.V.


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.


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.


Tojo M.,Global Nuclear Fuel Japan | Suzuki H.,Global Nuclear Fuel Japan | Sato H.,Global Nuclear Fuel Japan | Iwamoto T.,Global Nuclear Fuel Japan
Journal of Nuclear Science and Technology | Year: 2015

The source range monitors (SRMs) and the start-up range neutron monitors (SRNMs) are important instruments from the BWR criticality safety viewpoints. There is a limitation of the minimum count rate (3cps) to guarantee the normality of the SRMs/SRNMs. After the long outage, this limitation is critical for the fuel shuffling due to the decay of the neutron sources in the fuel. The neutron source intensity evaluation method based on a micro burn-up model and the predictor function of the SRM/SRNM count rate are developed in AETNA01, GNF's three-dimensional neutronic-thermal hydraulic boiling water reactor (BWR) core simulator. These new functions are validated through the comparisons between operating BWR's measured data after shutdown and during shuffling. Through these comparisons, high accuracy of the SRM/SRNM count rate predictor of AETNA01 was presented. © 2015 Atomic Energy Society of Japan. All rights reserved.


Yamamoto A.,Nagoya University | Iwata T.,Nagoya University | Iwata T.,Global Nuclear Fuel Japan | Yamane Y.,Nagoya University
Progress in Nuclear Energy | Year: 2011

Multicycle optimization was carried out by assuming power sharing of each fuel batch as an independent parameter; that is, the power sharing of each fuel batch was considered as an optimization variable. The steepest descent method was used to optimize the power sharing for multiple cycles. Two different optimizations were carried out, i.e., multicycle and successive single-cycle optimizations. In the former, the power sharing of each fuel batch in each cycle was simultaneously optimized for multiple cycles. In the latter, optimization of the power sharing in a cycle was carried out, and then optimization in the next cycle was carried out. Maximization of discharge burnup and minimization of the number of fresh fuel assemblies were considered as the objective functions. The calculation results qualitatively and quantitatively clarify the implicit adverse effect of the single-cycle optimization, which is usually used in current core designs. Under the calculation conditions of the present study, the difference in the number of fresh fuel assemblies between multicycle and successive single-cycle optimizations is 2-3 fuel assemblies per cycle. Comparison of the power sharing obtained by both methods would provide insights to correct the adverse effect of the single-cycle optimization. © 2011 Elsevier Ltd. All rights reserved.


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 | Hashizume K.,Kyushu University
Progress in Nuclear Energy | Year: 2012

To understand the basic oxidation kinetics of alloying elements which is considered to be strongly related with the corrosion and hydrogen pickup, the depth profiles of chemical states of alloying elements (Cr and Fe) were measured in the oxide layer of Zr-0.5Sn-1.0Cr-0.5Fe alloys. The depth profiles were obtained by combinations of a surface-sensitive XANES and an extremely low energy Ar ion sputtering. The XANES measurements revealed that the chemical states of alloying elements (Fe and Cr) varied with the depth in the oxide layer. Especially in the oxide layer formed in steam, a decrease of the fractions of oxidation states was significant rather than that in LiOH solution. In the oxide layer formed in steam, the oxidation rate of chromium was faster than iron by a factor of approximately 2. © 2011 Elsevier Ltd. All rights reserved.


Patent
Global Nuclear Fuel Japan | Date: 2011-06-29

A fuel assembly has a regular dodecagon fuel rod arrangement, in which a single fuel rod is provided to each apex of a regular dodecagon having lines with a length of a. Assuming that a direction within a horizontal plane is a transverse direction, and a direction perpendicular to the transverse direction is a longitudinal direction, regular dodecagon fuel rod arrangements are arranged in regular intervals in the transverse direction and the longitudinal direction. In the transverse direction, two adjacent regular dodecagon fuel rod arrangements are arranged so that the opposing two sides of the regular dodecagons are parallel to each other with the distance of ma (m : nonnegative integer) apart from each other. With respect to the longitudinal direction, two adjacent regular dodecagon fuel rod arrangements are arranged so that the opposing two sides of the regular dodecagons are parallel to each other with the distance of na (n : nonnegative integer) apart from each other.

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