Pressurized water reactor (PWR) vessel internals components can experience material aging and degradation due to irradiation . The Electric Power Research Institute (EPRI), under sponsorship of the Materials Reliability Program (MRP), is developing Reactor Internals Inspection and Evaluation (I&E) Guidelines mainly to support U.S. license renewed plants. These guidelines are organized around a framework and strategy,  and , for managing the effects of aging in PWR internals as shown in Figure 1, dependent on a substantial database of material data and supporting results. The key steps include the following: the development of screening criteria, with susceptibility levels for the eight postulated aging mechanisms relevant to reactor internals and their effects ; an initial component screening and categorization step, using the susceptibility levels to identify the relative susceptibility of the components; a functionality assessment of degradation for components and assemblies of components; and finally aging management strategy development combining the results of functionality assessment with component accessibility, operating experience, existing evaluations, and prior examination results to determine the appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections. The purpose of this functionality analysis is to provide a best estimate evaluation of the reactor internals core barrel assembly for materials degradation up to 60 years of operation. The stainless steel material model employed in the calculations is an irradiated material-specific constitutive model for use in a finite element analysis . The material model accounts for the effects of plasticity, irradiation assisted stress corrosion cracking (IASCC), irradiation creep-stress relaxation, void swelling, and embrittlement as a function of temperature and fluence  and . The study 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, detailed finite element models were developed capable of representing the complex interactions between the components. The goal of this study is to characterize the potential failure modes, spatial and chronological distribution of 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 major physics fields. 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 physics field involved, which facilitates the development of operating stress fields within the RV internals. The three major physics fields and their relations are illustrated in Figure 2. This paper focuses on the CFD/CHT aspects of the overall analysis for the B&W designed RV internals and provides information on the state-of-the-art multi-physics approach employed.
Materials Aging Degradation of Reactor Vessel Internals: Part I — Thermal Hydraulics Evaluation
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Liszkai, TR, Yee, NS, Smotrel, JR, & Demma, A. "Materials Aging Degradation of Reactor Vessel Internals: Part I — Thermal Hydraulics Evaluation." Proceedings of the ASME 2008 Pressure Vessels and Piping Conference. Volume 4: Fluid-Structure Interaction. Chicago, Illinois, USA. July 27–31, 2008. pp. 303-312. ASME. https://doi.org/10.1115/PVP2008-61804
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