Fatigue crack growth behavior of pressure vessel steels and submerged arc weldments in a high-temperature pressurized wa
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I.
INTRODUCTION
W H I L E the termination of the life of a structure or component may be based on the critical flaw size for large-scale rupture as calculated from the material's fracture toughness, it must be recognized that the total useful life of a cyclically loaded component such as a reactor pressure vessel is dependent upon the rate of growth of flaws from a subcritical size to a critical size. Therefore, an understanding of the crack growth characteristics of the required reactor pressure vessel-grade materials under appropriate service conditions is essential to evaluate the useful life of a pressure vessel. Moreover, the corrosive environment F C G R data initially included in Section XI of the ASME Boiler and Pressure Vessel Code, relative to pressure retaining materials for vessels utilized in nuclear applications, was both sparse and questionable and, in addition, was directly applicable only to ferritic pressure vessel steels with minimum yield strengths of 345 MPa (50 ksi). m This lack of corrosive environment FCGR data for those pressure vessel steels utilized in nuclear reactors was most likely due to the great degree of difficulty typically experienced when generating this data. Per this investigation, the effects of an HPW environment on the fatigue crack growth behavior of SA508 2a and SA533 Gr A C1 2 pressure vessel steels and the corresponding automatic submerged arc weldments were evaluated at load ratios (R = Pmin/P...... where Pmin and Pmax are the applied minimum and maximum loads, respectively) of 0.20 and 0.50. These F C G R tests were initiated as part of a combined French-Westinghouse fourparty [Commissariat a l'Energie Atomique (CEA), Elec-
P.K. LIAW, J.A. BEGLEY, Fellow Engineers, and W.A. LOGSDON, Senior Engineer, are with the Westinghouse Research and Development Center, 1310 Beulah Road, Pittsburgh, PA 15235. Manuscript submitted October 6, 1988. METALLURGICAL TRANSACTIONS A
tricite de France (EdF), Framatome (FRAMATOME), and Westinghouse Electric Research and Engineering for Atomic Systems, Inc. (WEREAS)] joint research and development program. In addition, a mechanistic understanding of fatigue crack growth kinetics in the HPW environment was provided by extensive fractographic and sulfur-inclusion examinations. Base line 24 ~ and 288 ~ (75 ~ and 550 ~ air environment f21 and primary side pressurized water reactor (PWR) environment I31 FCGR tests were previously conducted on the SA508 C1 2a and SA533 Gr A C1 2 base materials and the corresponding automatic submerged arc weldments, as well as on two similar base materials, SA508 C1 3a and SA533 Gr B C1 2. The near-threshold FCGR properties of all four base materials, as well as the two automatic submerged arc weldments, were additionally generated in an air environment at 24 ~ and 288 ~ (75 ~ and 550 ~ at load ratios of 0.20 and either 0 . 5 0 o r 0 . 7 0 . [41
The primary and secondary flow loops of pressurized water reactors (PWR) have different water chemistries. Primary water contains a boric acid addition, which serves
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