Materials Aging Degradation of Reactor Vessel Internals: Part II—Structural Evaluation and Global Finite Element Models

Abstract

Pressurized water reactor (PWR) vessel internals components can experience material aging and degradation due to irradiation [1]. The Electric Power Research Institute (EPRI), under sponsorship of the Materials Reliability Program (MRP), developed PWR Internals Inspection and Evaluation (I&E) Guidelines mainly to support license renewal of U.S. plants [2]. The functionality analysis of reactor internals components and assemblies was one of the tools used to develop these guidelines. The purpose of the functionality analysis is to provide a best estimate evaluation of the reactor internals core barrel assembly for materials degradation and to assess whether the components retain their function. The evaluation uses an irradiated material-specific constitutive model for use in a finite element analysis representing the current state of knowledge for plasticity, creep, stress relaxation, void swelling, and embrittlement [3], 4, [5]. This constitutive model is a function of temperature and fluence. The analysis focuses on finding the integrated effects of material aging combined with steady-state operational characteristics of the reactor vessel (RV) internals. In order to evaluate the potential failure mechanisms of the core barrel assembly, finite element models were developed capable of representing the complex interactions between the components. The goal of this specific analysis is to characterize the potential failure modes, spatial and chronological distribution of potential component failures for a representative model of the Babcock & Wilcox (B&W) designed plants. Evaluation of the reactor vessel internals for materials aging degradation involves three analytical calculations. Radiation calculations of the core provide essential information on radiation dose and heat rates, due to gamma-heating, of the RV internals. The computational fluid dynamics domain (CFD) allows the evaluation of the RV internals temperatures through conjugate heat transfer (CHT) analysis coupled with coolant flow. Detailed structural analysis of the RV internals components and bolted connections is the third major analytical calculation, which facilitates the development of operating stress fields within the RV internals. Structural analysis was performed as two parts. First, a global structural model of the core barrel assembly was used to represent the interaction of components of the core barrel assembly during 60 years of operation. The global model does not include detail of the areas of stress concentration within bolted connections, therefore local models of selected bolts were developed. Results of both the global and local models were used as a basis for evaluating age-related effects. The description of the functionality analysis for the B&W designed RV internals is divided into three papers. Part I was presented in PVP-2008 [6] and included a description of the overall methodology with special attention to CFD-CHT evaluations. Part II, detailed in this paper, describes global structural finite element models. Part III, to be also presented at PVP-2009 [7], presents a description of local models of bolted connections, results, and conclusions.