Microstructure-based Numerical Simulation to Predict Mechanical Properties of 316CW Stainless Steel H-J. Joo1, Y-S. Chang2,∗ 1 Graduate School, Kyung Hee University, Yongin 17104, Republic of Korea 2 Department of Nuclear Engineering, Kyung Hee University, Yongin 17104, Republic of Korea ∗ yschang@khu.ac.kr Keywords: 316CW Steel, Finite Element Analysis, Irradiation Effect, Nano-indentation, Reactor Vessel Internals Although most of reactor vessel internals (RVIs) have been designed by austenite stainless steels with high strength and ductility, not only possibility of material property changes due to harsh environments but also a few failures caused by microstructural defects et cetera were reported in the RVIs of operating nuclear power reactors. Particularly, the investigation on irradiated materials is important but direct measurement of their properties is not easy because of complexities associated with preparation of neutron irradiated standard specimens and test facilities. Therefore, recently, several researches were carried out to incorporate irradiation embrittlement depending on dose levels [1] by establishing alternative ion irradiation conditions and using miniature specimens. In this study, effectiveness of a microstructure-based simulation method was examined through the prediction of as-received and irradiated mechanical properties of 316CW (Cold Worked) stainless steel. At first, dislocation density-related material constitutive parameters were determined by comparing nano-indentation test data [2] and corresponding numerical analysis results. Then, microscopic stress-strain curves were derived based on the calibrated constitutive parameters and converted into macroscopic stress-strain curves in use of existing correlations [3]. Finally, influence of different dose levels was assessed via a series of FEA and the subsequent procedure, of which details and key findings will be discussed. References [1] Edwards, D.J., Simonen E.P., Garner F.A., Greenwood L.R., Oliver B.M., and Bruemmer S.M., Influence of irradiation temperature and dose gradients on the microstructural evolution in neutron-irradiated 316SS, Journal of Nuclear Materials, Vol. 317, pp. 32-45, 2003. [2] Pokor, C., Brechet, Y., Dubuisson, P., Massoud, J. P., and Barbu, A., Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the Microstructure, Journal of Nuclear Materials, Vol. 326 pp. 19-29, 2004. [3] Patel, D. K. and Kalidindi, S. R., Correlation of spherical nanoindentation stress-strain curves to simple compression stress-strain curves for elastic-plastic isotropic materials using finite element models, Acta materialia, Vol. 112, pp. 295-302, 2016. 56
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