Abstract
The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.
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References
K. Takakura et al., “CRACK GROWTH BEHAVIOR OF NEUTRON IRRADIATED L-GRADE AUSTENITIC STAINLESS STEELS IN SIMULATED BWR CONDITIONS”, Proc. 14th Int. Conf. Environmental Degradation of Materials in Nuclear Power Systems, 1192–1203 (2009).
Y. Kaji et al., “INFLUENCE OF MICROSTRUCTURE ON IASCC GROWTH BEHAVIOR OF NEUTRON IRRADIATED TYPE 304 AUSTENITIC STAINLESS TEELS IN SIMULATED BWR CONDITION”, Proc. 14th Int. Conf. Environmental Degradation of Materials in Nuclear Power Systems, 1181–1191 (2009).
T.M. Karlsen, P. Bennett, and N.W. Høgberg, “In-Core Crack Growth Rate Studies on Irradiated Austenitic Stainless Steels in BWR and PWR Conditions in the Halden Reactor”, Proc. 12th Int. Conf. Environmental Degradation of Materials in Nuclear Power System — Water Reactors -, TMS, 337–348 (2005).
J. Nakano, T.M. Karlsen, and N.W. Høgberg, “Present State Of Crack Growth And Crack Initiation Studies Performed In The Halden Reactor”, Proc. 15th Int. Conf. Nuclear Engineering, ICONE 15–10399 in CD-ROM (2007).
Y. Kaji et al., “COMPARISON OF SCC GROWTH RATE BETWEEN IN-CORE AND EX-CORE TESTS IN BWR SIMULATED HIGH TEMPERTURE WATER”, Proc. 13th Int. Conf. Environmental Degradation of Materials in Nuclear Power Systems, in CD-ROM (2007).
Y. Kaji et al., “In-Core SCC Growth Behavior of Type 304 Stainless Steel in BWR Simulated High-Temperature Water at JMTR”, J. Nucl. Sci. Technol. 45, 725–734 (2008).
O.K. Chopra, E.E. Gruber, and W.J. Shack, “Crack Growth Behavior of Irradiated Austenitic Stainless Steels in High-Purity Water at 289°C”, Proc. 11th Int. Conf. Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, in CD-ROM (2003).
S. Ooki et al., “STUDY ON SCC GROWTH BEHAVIOR OF BWR CORE SHROUD”, Proc. 12th Int. Conf. Environmental Degradation of Materials in Nuclear Power System — Water Reactors -, TMS, 365–376 (2005).
JANTI, “BWR Core Internals Inspection and Evaluation Guideline — Core Shroud — (4th Edition)”, JANTI-VIP-05–4th Edition (2008).
JSME, “Codes for Nuclear Power Generation Facilities — Rules —”, JSME S NA1–2008 (2008).
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© 2011 TMS (The Minerals, Metals & Materials Society)
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Chimi, Y., Kasahara, S., Ise, H., Kawaguchi, Y., Nakano, J., Nishiyama, Y. (2011). In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR. In: Busby, J.T., Ilevbare, G., Andresen, P.L. (eds) Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors. Springer, Cham. https://doi.org/10.1007/978-3-319-48760-1_74
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DOI: https://doi.org/10.1007/978-3-319-48760-1_74
Publisher Name: Springer, Cham
Online ISBN: 978-3-319-48760-1
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