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Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji
Journal of Nuclear Materials, 505, p.44 - 53, 2018/07
Times Cited Count：2 Percentile：20.93(Materials Science, Multidisciplinary)A calculation model was constructed to systematically study the effects of environmental conditions (i.e. Cr concentration in sodium, test temperature, axial temperature gradient of fuel pin, and sodium flow velocity) on Cr dissolution behavior. Chromium dissolution was largely influenced by small changes in Cr concentration (i.e. chemical potential of Cr) in liquid sodium in the model calculation. Chromium concentration in sodium coolant, therefore, should be recognized as a critical parameter for the prediction and management of Cr dissolution behavior in the sodium-cooled fast reactor (SFR) core. Because the fuel column length showed no impact on dissolution behavior in the model calculation, no significant downstream effects possibly take place in the SFR fuel cladding tube due to the much shorter length compared with sodium loops in the SFR plant and the large axial temperature gradient. The calculated profile of Cr concentration along the wall-thickness direction was consistent with that measured in BOR-60 irradiation test where Cr concentration in sodium bulk flow was set at 0.07 wt ppm in the calculation.
Kikuchi, Kenji; Kurata, Yuji; Saito, Shigeru; Futakawa, Masatoshi; Sasa, Toshinobu; Oigawa, Hiroyuki; Wakai, Eiichi; Miura, Kuniaki*
Journal of Nuclear Materials, 318(1-3), p.348 - 354, 2003/05
Times Cited Count：28 Percentile：84.98(Materials Science, Multidisciplinary)Corrosion test of austenitic stainless tube was done under the flowing Pb-Bi condition during 3000 hrs at 450
C. Specimen is 316SS produced as a tubing form with 13.8 mm outer diameter, 2 mm thickness and 40 cm length. During the operation, maximum temperature, temperature difference and flow velocity of Pb-Bi at the specimen were kept at 450C, 50C, and 1m/s, respectively. After the test, specimen and components of the loop were cut and examined by optical microscope, SEM, EDX, WDX and X-ray diffraction. Pb-Bi adhered on the surface of the specimen even after Pb-Bi was drained out to the storage tank from the circulating loop. Different results from a stagnant corrosion test were that the specimen surface became rough and the corrosion rate was maximally 0.1mm/3000hrs. And mass transfer from the high temperature to the lower temperature area was observed: crystals of Fe-Cr were found on the tube surface in low-temperature part. The size of crystal was 0.1 
0.2 mm. The depositing crystal was ferrite grain and the chemical composition ratio (mass%) of Fe to Cr was 9:1.
O and LiAlO spheres during the VOM-22H experimentKurasawa, Toshimasa; Hollenberg, G. W.*; Watanabe, H.
Proc.Int.Symp.on Fusion Reactor Blanket and Fuel Cycle Technology, p.43 - 46, 1987/00
no abstracts in English
; Konishi, Satoshi; ; Kurasawa, T.; ; Naruse, Yuji
Journal of Nuclear Materials, 122-123, P. 934, 1984/00
no abstracts in English
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Uwaba, Tomoyuki; Kaito, Takeji; Furukawa, Tomohiro; Kato, Shoichi
no journal, ,
Maximum temperature of ODS steel cladding tube for long life fast reactor fuel is very high (approximately 700C) in normal operation condition. It was reported that, in reactor operation, mass transfer phenomena (dissolution, deposition, penetration) took place as a result of increased solubility of steel constituent elements in liquid Na. The driving force of these phenomena is the chemical potential gap of solute elements in steel and liquid Na, which is dependent of not only temperature but also other factors such as impurity concentrations in Liquid Na. For appropriately evaluating experimental data and predicting the corrosion behavior in actual plant, it is required to list up the key factors including other factors than temperature and residence time and understand the effects of these factors. In this study, transfer behavior of Cr (main alloying element of ODS steel) is discussed; modelling and numerical calculation were carried out on Cr dissolution behavior from fast reactor fuel cladding tube into liquid Na.
