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It is commonly believed that the fracture toughness of a brittle material can be characterized by a single parameter such as the stress intensity factor. In this study, it was demonstrated that when the crack is highly constrained, the first nonsingular opening stress term at the crack tip, in addition to the Kfield (the singular stress term), is necessary to predict fracture. Fracture experiments were conducted using plexiglass specimens with a center crack. Relatively rigid metallic end tabs were used to generate boundary constraints on the specimen. The level of constraint was varied by varying the gage length between the end tabs. For a given crack length, the fracture load increases as the gage length decreases. If the stress intensity factor is used to determine the corresponding fracture toughness of plexiglass, the experimental data would indicate that the fracture toughness decreases as the gage length decreases. This is equivalent to saying that the fracture toughness of a brittle material can be affected by boundary conditions. It was shown that this behavior is the result of a diminishing size of the K -dominance zone and that the stress intensity factor alone cannot fully capture the fracture force. A new constant parameter was introduced to account for the effect of the near-tip nonsingular stress field on fracture.

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Leak-Before-Break (LBB) analysis allows nuclear power plants to eliminate consideration of the dynamic effects of pipe rupture from the plant design basis for the affected Class I piping system, and remove protective hardware such as pipe whip restraints, jet impingement barriers, etc. Accurately calculating leakage rate of the postulated circumferential flaw in the pipe line system is critical for the LBB analysis. To obtain the final leakage rate, two calculations have to be performed: critical circumferential flaw size and then, the corresponding leakage rate through the flaw. If present, including crack face pressure in the evaluation will help enlarge the crack opening distance and hence, increase the leakage flow rate per unit flaw length. However, it creates larger force and bending moment on the crack, which reduces the critical flaw size and therefore, limits the final leakage amount. The opposite effect of crack face pressure on critical flaw size and leakage rate per unit flaw size needs to be evaluated ascertain its overall effect on LBB evaluation. In this paper, leakage rates are calculated and compared for cases with or without consideration of crack face pressure for various piping systems with typical dimensions and loadings. The crack face pressure effect on the final leakage rate of LBB evaluation is discussed.

Prolonged exposure of cast austenitic stainless steels (CASS) to reactor coolant operating temperatures has been shown to lead to some degree of thermal aging embrittlement (reduction in fracture toughness of the material as a function of time). The fracture toughness data for the most severely aged CASS materials were found to be similar to those reported for some austenitic stainless steel weld metal, in particular weld metal from submerged arc welds (SAW). Such similarity offers the possibility for applying periodic inservice inspection flaw acceptance criteria, currently referenced in the ASME Code Section XI, Subsection IWB, for SAW and shielded metal arc weld (SMAW), to CASS component inservice inspection results. This paper presents the data to support both the proposed screening criteria (based on J-R crack growth resistance) for evaluation of the potential significance of the effects of thermal aging embrittlement for Class 1 reactor coolant system and primary pressure boundary CASS components, for those situations where the effects of thermal aging embrittlement are found to be potentially significant. The fitness for continued service is based on the comparison of the limiting fracture toughness data for Type 316 SAW welds and the lower-bound fracture toughness data reported for high-molybdenum, high delta-ferrite, statically and centrifugally-cast CASS materials. These comparisons and the associated flaw acceptance criteria can be used to justify exemptions from current ASME Code Section XI inservice inspection requirements through flaw tolerance evaluation (e.g., see ASME Nuclear Code Case N-481).