[1]
U. S. Nuclear Regulatory Commission, 10 CFR 50 Appendix G, Fracture Toughness Requirements.
Google Scholar
[2]
Lee, T. J., Choi, J. B., Kim, Y. J., & Park, Y. W.. A parametric study on pressure–temperature limit curve using 3-D finite element analyses. Nuclear engineering and design, (2002), 214(1-2), 73-81.
DOI: 10.1016/s0029-5493(02)00016-x
Google Scholar
[3]
Pennell, W. E., Malik, S. N. M. Structural integrity assessment of aging nuclear reactor pressure vessels. Nuclear engineering and design, (1997), 172(1-2), 27-47.
DOI: 10.1016/s0029-5493(96)00006-4
Google Scholar
[4]
González-Albuixech, V. F., Qian, G., & Niffenegger, M. Integrity analysis of reactor pressure vessels subjected to pressurized thermal shocks by XFEM. Nuclear Engineering and Design, (2014), 275, 336-343.
DOI: 10.1016/j.nucengdes.2014.04.038
Google Scholar
[5]
Kharchenko, V. V., Kobel'skii, S. V., Kravchenko, V. I., Chirkov, A. Y., & Zvyagintseva, A. A. Determination of stress intensity factors for semielliptical surface cracks in the WWER-1000 reactor pressure vessel from the results of solving thermoelasticity boundary value problems based on the mixed mesh-projection scheme of the finite element method. Strength of materials, (2007).
DOI: 10.1007/s11223-007-0019-z
Google Scholar
[6]
Jhung, M. J., Kim, S. H., Choi, Y. H., Jung, S., Kim, J. M., Kim, J. H., ... & Kang, K. S. Structural integrity assessment of reactor pressure vessels during pressurized thermal shock. Journal of mechanical science and technology, (2008).
DOI: 10.1007/s12206-008-0421-x
Google Scholar
[7]
Qian, G., Niffenegger, M. Procedures, methods and computer codes for the probabilistic assessment of reactor pressure vessels subjected to pressurized thermal shocks. Nuclear Engineering and Design, (2013), 258, 35-50.
DOI: 10.1016/j.nucengdes.2013.01.030
Google Scholar
[8]
Qian, G., Niffenegger, M. Deterministic and probabilistic analysis of a reactor pressure vessel subjected to pressurized thermal shocks. Nuclear Engineering and Design, (2014), 273, 381-395.
DOI: 10.1016/j.nucengdes.2014.03.032
Google Scholar
[9]
ASME Boiler and Pressure Vessel Code Sec. XI, Protection against non-ductile failure, Appendix G. (2010).
Google Scholar
[10]
U. S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plant: LWR Edition (Formerly issued as NUREG-75/057), Revision 1, Office of Nuclear Reactor Regulation Washington D. C., July (1981).
DOI: 10.2172/211545
Google Scholar
[11]
WRC, PVRC Ad Hoc Group on Toughness Requirements, PVRC recommendations on toughness requirements for ferritic materials. WRC bulletin 175. (1972).
Google Scholar
[12]
Hernández-Gómez, L. H., Pérez-Escobar, B. L., Beltrán-Fernández, J. A., Flores-Campos, J. A., Pérez-Montejo, S., Jiménez-Santiago, K. I., & Urriolagoitia-Calderón, G. M. (2016).
DOI: 10.4028/www.scientific.net/ddf.370.162
Google Scholar
[13]
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99. Radiation embrittlement of reactor vessel materials,. Revision 2, (1988).
Google Scholar
[14]
U. S. Nuclear Regulatory Commission, 10 CFR 50 Appendix A, General Design Criteria for Nuclear Power Plants,. Paragraph 31.
Google Scholar
[15]
ANSYS, Inc. ANSYS Mechanical User´s Guide (2013).
Google Scholar
