TY - GEN
T1 - Evaluation of weight loss of stainless steels in supercritical water
AU - Maruno, Yusaku
AU - Kaneda, Junya
AU - Kasahara, Shigeki
AU - Saito, Norihisa
AU - Shikama, Tatsuo
AU - Matsui, Hideki
PY - 2009/1/1
Y1 - 2009/1/1
N2 - The supercritical water-cooled reactor (SCWR) is one of the Generation IV nuclear systems. According to a planned design for the SCWR, the temperature of the main steam is supposed to be 560°C and the surface temperature of the fuel cladding, which has a thickness of only 0.45 mm, is supposed to reach 700°C. General corrosion is one of the most critical issues for designing the fuel cladding when taking these temperature conditions into account. Therefore, the general corrosive behavior of candidate cladding materials under the SCWR conditions needs to be evaluated. In particular, from the viewpoint of a safe-life design for the SCWR, it is necessary to evaluate the amount of weight loss, which reflects the corrosion resistance, under the SCWR conditions before selecting the most promising materials for the fuel claddings. This paper reports on the weight loss of Type304, Type316L, Type310S, 12Cr-1Mo-1WVNb, modified Type316L, and modified Type310S under 500, 550, and 600°C supercritical water conditions. After the corrosion tests were finished, the oxide film from the specimen was removed using molten lithium. The weight loss was measured as the difference in weight of the same specimen before and after the corrosion test without the oxide film. The results showed that the weight loss increased along with the test temperature. On the basis of the Arrhenius equation, a linear relationship was observed between the weight loss and the reciprocal of the temperature. Type310S and modified Type310S had a smaller weight loss than Type316L, Type304, and 12Cr-1Mo-1WVNb. In addition, the weight loss of these candidate materials at 700°C during 50000h was extrapolated by the Arrhenius equation. Meanwhile, the reduced thicknesses of Type310S and modified Type310S were estimated as 45 μm or less, which is 10% or less of the thickness of the fuel claddings. Based on the above results, Type310S and modified Type310S were the most promising alloys for the fuel cladding.
AB - The supercritical water-cooled reactor (SCWR) is one of the Generation IV nuclear systems. According to a planned design for the SCWR, the temperature of the main steam is supposed to be 560°C and the surface temperature of the fuel cladding, which has a thickness of only 0.45 mm, is supposed to reach 700°C. General corrosion is one of the most critical issues for designing the fuel cladding when taking these temperature conditions into account. Therefore, the general corrosive behavior of candidate cladding materials under the SCWR conditions needs to be evaluated. In particular, from the viewpoint of a safe-life design for the SCWR, it is necessary to evaluate the amount of weight loss, which reflects the corrosion resistance, under the SCWR conditions before selecting the most promising materials for the fuel claddings. This paper reports on the weight loss of Type304, Type316L, Type310S, 12Cr-1Mo-1WVNb, modified Type316L, and modified Type310S under 500, 550, and 600°C supercritical water conditions. After the corrosion tests were finished, the oxide film from the specimen was removed using molten lithium. The weight loss was measured as the difference in weight of the same specimen before and after the corrosion test without the oxide film. The results showed that the weight loss increased along with the test temperature. On the basis of the Arrhenius equation, a linear relationship was observed between the weight loss and the reciprocal of the temperature. Type310S and modified Type310S had a smaller weight loss than Type316L, Type304, and 12Cr-1Mo-1WVNb. In addition, the weight loss of these candidate materials at 700°C during 50000h was extrapolated by the Arrhenius equation. Meanwhile, the reduced thicknesses of Type310S and modified Type310S were estimated as 45 μm or less, which is 10% or less of the thickness of the fuel claddings. Based on the above results, Type310S and modified Type310S were the most promising alloys for the fuel cladding.
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M3 - Conference contribution
AN - SCOPUS:84907922830
T3 - International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009
SP - 2267
EP - 2276
BT - International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009
PB - Atomic Energy Society of Japan
T2 - International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009
Y2 - 10 May 2009 through 14 May 2009
ER -