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This paper proposes computationally efficient frequency domain formulations for two well‐known multiaxial fatigue failure criteria, namely Matake’s critical plane criterion and Crossland’s criterion. For that purpose, it is shown how fatigue‐related variables involved in both criteria can be estimated from the power spectral density matrix of the local stress vector. The finite element model of an example structure is then used to illustrate the application of the proposed frequency domain approaches. It is observed that both frequency domain formulations produce consistent results when compared with those obtained in the time domain from Monte‐Carlo simulations of local stress vectors while offering tremendous computer savings. A frequency domain tool indicating whether the principal stress directions do rotate with time or not during the loading at a given location in the structure is also presented.

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Full structural weld overlays (FSWOLs) have been used extensively as a repair/mitigation technique for primary water stress corrosion cracking (PWSCC) in pressurizer nozzle dissimilar metal (DM) welds. To support an approved FSWOL design and safety submission for British Energy pressurized water reactor (PWR) nozzles, an in-depth evaluation was performed to assess the effects of a FSWOL on the through-wall residual stress distribution in safety/relief pressurizer nozzles. Two safety/relief pressurizer nozzle mockups were fabricated based on British Energy’s PWR nozzle design. One mockup included the nozzle to safe-end DM weld and the safe-end to stainless steel weld while the second mockup included the DM weld, the stainless steel weld and a Westinghouse-designed structural weld overlay. The mockups were fabricated utilizing materials and techniques that represented the plant-specific nozzles as closely as possible and detailed welding parameters were recorded during fabrication. All welds were subsequently nondestructively evaluated (NDE). A thorough review of the detailed fabrication records and the NDE results was performed and several circumferential positions were selected on each mockup for subsequent residual stress measurement. The through-wall residual stress profiles were experimentally measured through the DM weld centerline at the selected circumferential positions using both the deep hole drilling (DHD) and incremental deep hole drilling (iDHD) measurement techniques. In addition to experimental residual stress measurements, the through-wall residual stress profiles were simulated using a 2-D axisymmetric ANSYS™ finite element (FE) model. The model utilized kinematic strain hardening and the temperature constraint method which greatly simplified the simulation as compared to detailed heat source modeling methods. A range of residual weld stress profiles was calculated by varying the time at which the temperature constraints were applied to the model. The simulation results were compared to the measurement results. It was found that the effects of the FSWOL were principally three fold. Specifically, the FSWOL causes a much deeper compressive stress field, i.e., the overlay shifts tension out towards the outside diameter surface. Further, the FSWOL reduces tension in the underlying dissimilar metal weld, and finally, the FSWOL causes higher peak compressive and tensile residual stresses, both of which move deeper into the nozzle wall after the overlay is applied. Relatively good agreement was observed between the FE results and the measurements results.

Plants in the United States (U.S.) and many plants outside of the U.S. are required to meet the regulations of the Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61. The Alternate Pressurized Thermal Shock (PTS) Rule (10 CFR 50.61a) was approved by the U.S. Nuclear Regulatory Commission (NRC) and included in the Federal Register, with an effective date of February 3, 2010. This Alternate Rule provides a new metric and screening criteria for PTS. This metric, RTMAX-X, and the corresponding screening criteria are far less restrictive than the RTPTS metrics and screening criteria in the original PTS Rule (10 CFR 50.61). The Alternate PTS Rule was developed through probabilistic fracture mechanics (PFM) evaluations performed for selected U.S. pilot plants. A Generalization Study was also performed which determined that the plants used for these evaluations were representative of and applicable to the U.S. Pressurized Water Reactor (PWR) nuclear power plant fleet. Plants outside of the U.S. may be interested in implementing the Alternate PTS Rule. However, direct implementation of the Alternate PTS Rule may not be possible due to differences in plant design, embrittlement prediction techniques, inservice inspection requirements, etc. The objective of this paper is to explore the use the Alternate PTS Rule by PWR plants outside of the U.S. by proposing methods to account for the potential differences mentioned above.