Full-scale 3-D finite element modeling of a two-loop pressurized water reactor for heat transfer, thermal–mechanical cyclic stress analysis, and environmental fatigue life estimation

Abstract

This paper discusses a system-level finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent sequentially coupled thermal-mechanical stress analysis were performed for typical thermalmechanical fatigue cycles. The in-air fatigue lives of example components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in US-NRC report: NUREG-6909.

Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

“…We developed finite element (FE) models for both heat transfer analysis and for subsequent thermal-mechanical stress analysis. In our earlier work [11,12] we presented a preliminary/skeletal FE model of an overall reactor consisting of a reactor pressure vessel, hot leg, cold leg, and steam generator. In this work, we present a detailed FE model of the reactor pressure vessel and its nozzle only.…”

“…For the requirement of modeling cracks, 3D models were developed and meshed by using eight nodded 3D brick elements. In our previous work [11,12] we found that eight-node linear elements (DC3D8) were sufficient to model heat transfer compared to a computationally expensive counterpart of 20-node brick elements (DC3D10). For stress analysis, the corresponding C3D8, 8-node linear elements were used.…”

“…For stress analysis, the corresponding C3D8, 8-node linear elements were used. Note that in our earlier work [11,12], we considered other components such as the HL, CL, and steam generator; in the present work, however, we have not considered those components to reduce the computational burden by limiting the number of finite elements. Instead, we included detailed geometry of HL/CL nozzles and increased the number of elements along the thickness direction to allow modeling of preexisting cracks.…”

“…Although most modern nuclear plants are designed to follow grid demands to a certain extent through consideration of a large safety factor, no study (at least none reported in the open literature) has been undertaken to determine the stress-strain state of reactor components under grid-load following. Most previous research on NPP component safety assessment, including our own [11][12], is based on a stress analysis of components using simplified design transients. For accurate structural integrity assessment of NPP components, it is necessary to perform structural fatigue evaluation under more realistic loads [13][14][15].…”

“…Argonne National Laboratory (ANL), under the sponsorship of the Department of Energy's Light Water Reactor Sustainability (LWRS) program, has been involved with extensive material testing [18,19] and mechanistic modeling [11,12,32] for assessing the structural integrity of NPP components under design and extended service conditions. In this paper we present detailed results from thermal mechanical stress analysis of a reactor pressure vessel (RPV) and its nozzle (both with and without preexisting crack) under typical reactor heat-up, cool-down, and load-following modes.…”