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and Zr at precisely controlled high temperaturesShirasu, Noriko; Sato, Takumi; Suzuki, Akihiro*; Nagae, Yuji; Kurata, Masaki
Journal of Nuclear Science and Technology, 60(6), p.697 - 714, 2023/06
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)Interaction tests between UO and Zr were performed at precisely controlled high temperatures between 1840 and 2000 
C to understand the interaction mechanism in detail. A Zr rod was inserted in a UO crucible and then heat-treated at a fixed temperature in Ar-gas flow for 10 min. After heating in the range of 1890 to 1930 C, the Zr rod was deformed to a round shape, in which the post-analysis detected the significant diffusion of U into the Zr region and the formation of a dominant 
-Zr(O) matrix and a small amount of U-Zr-O precipitates. The abrupt progress of liquefaction was observed in the sample heated at around 1940 C or higher. The higher oxygen concentration in the -Zr(O) matrix suppressed the liquefaction progress, due to the variation in the equilibrium state. The U-Zr-O melt formation progressed by the selective dissolution of Zr from the matrix, and the selective diffusion of U could occur via the U-Zr-O melt.
Shirasu, Noriko; Suzuki, Akihiro*; Nagae, Yuji; Kurata, Masaki
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05
High temperature interaction tests between UO and Zr were performed at around 2173 K, to make clear the UO/ -Zr(O) interaction and the mechanism of degradation, for developing the improved models for advanced severe accident analysis codes. A Zr plate was inserted in a UO crucible, and heat treated at 2173 K in stream of Ar. After the heat-treatment, the samples were subjected to surface microanalysis. The middle region of Zr sample shows streak-like structures which are extended towered the top. It is confirmed that the streak-like structures were mainly consist of U from the EDX results, and the structures revealed that the U-rich phase was liquid during the heat-treatment. It seems that the U-rich liquid grew selectively toward the area where the oxygen concentration was low.
Hidaka, Akihide; Kudo, Tamotsu; Nakamura, Takehiko; Uetsuka, Hiroshi
Journal of Nuclear Science and Technology, 39(3), p.273 - 275, 2002/03
Times Cited Count：10 Percentile：54.86(Nuclear Science & Technology)no abstracts in English
JNC TN9400 2000-038, 98 Pages, 2000/04
JNC-TN9400-2000-038.pdf:7.49MB
As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.
; ;
JNC TN9400 2000-029, 87 Pages, 1999/11
JNC-TN9400-2000-029.pdf:5.11MB
The second Power-To-Melt (PTM) test, PTM-2, was performed in the experimental fast reactor "JOYO". AIl of the twenty-four fuel pins of the irradiation vehicle, B5D-2, for the PTM-2 test, were provided for post-irradiation-experiment (PIE) to evaluate the PTM values. ln this study, the PIE technique for PTM test was established and the PTM results were evaluated. The findings are as follows: (1) The maximum fuel-melting ratio on the transverse section was 10.7%, and was within the limit of fuel-melting in this PTM test enough. Unexpected fuel-melting amount to a ratio of 11.8% was found at 
24 mm below the peak power elevation in a test fuel pin, lt is possible that this arose from secondary fuel-melting. (2) Combination of metallographical observation with X-ray microanalysis of plutonium distribution was very effective for the identification of once-molten fuel zone. (3) The PTM evaluation suggested that dependence of the PTM on the fuel pellet density was stronger than that of previous foreign PTM tests, while the dependence on the pellet-cladding gap and the oxygen-to-metal ratio was indistinctly. The dependence on the cladding temperature and the fill gas composition was not shown as well.
Hidaka, Akihide; Nakamura, Jinichi; Sugimoto, Jun
Nucl. Eng. Des., 168(1-3), p.361 - 371, 1997/00
Times Cited Count：2 Percentile：23.04(Nuclear Science & Technology)no abstracts in English
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Journal of Nuclear Science and Technology, 24(1), p.12 - 22, 1987/01
Times Cited Count：1 Percentile：19.35(Nuclear Science & Technology)no abstracts in English
JAERI-M 4453, 21 Pages, 1971/05
no abstracts in English
and ZrShirasu, Noriko; Suzuki, Akihiro*; Nagae, Yuji; Kurata, Masaki
no journal, ,
The high temperature interaction tests between UO and Zr were performed at around 2173 K in stream of Ar, to make clear the UO and Zircaloy cladding interaction and the mechanism of degradation. After the heat-treatment, the samples were subjected to surface microanalysis. The mechanism of fuel degradation was more clarified.
Chai, P.; Yamashita, Susumu; Nagae, Yuji; Kurata, Masaki
no journal, ,
Various numerical simulations were performed by JUPITER code in order to validate the reliability of its multi-physics models, which were developed for evaluating the melting and relocation behavior of the core materials. By comparing with the previous experimental results, we could conclude that JUPITER code is a useful tool on severe accident analysis.
