During postulated high-pressure core melt accidents, high temperatures can be reached in the reactor coolant system (RCS) pressure boundary due to hot gases from the core melt process, which may lead to failure of the pressure boundary in terms of a leak or a break. Location, time and mode of the failure can have large implications on the further accident progression. Nevertheless, not for all possibly affected components and failure modes suitable assessment methods, in-depth structure mechanical analyses or in some cases even simple models for the integration into severe accident codes are available.This paper concentrates on assessment methods for gross boundary failure, e.g. of RCS piping. First, failure modes of components as well as the deformation and fracture behaviour of typical RCS materials in the severe accident loading regime are discussed. Established assessment methods based on linear damage accumulation and the use of Finite Element (FE) codes as well as a newly developed method based on a reduction to a 0-dimensional element is presented. The methods are validated and compared on a largescale experiment of a DN700 pipe and international benchmark analyses within the OECD/NEA activity COSSAL (Components and Structures under Severe Accident Loading). Results of the application to a representative severe accident scenario in a PWR and parametric studies are summarized.
Methods to demonstrate break preclusion for pressure retaining components in nuclear power plants are summarized in the newly developed German safety standard KTA 3206, especially the requirements for the leak-before-break assessment of crack-like leaks. In this context numerical and simplified calculation methods for both the leak opening area and fluid flow rates of crack-like leaks were examined and validated on selected experiments. Second topic of the paper is an accident analysis for a postulated leak in form of a through-wall crack in the pressurizer surge line (SL) of a German PWR Konvoi type. The initial size of the leak as well as the changing size depending on system pressure and temperature during the postulated leak accident was calculated with a FEM analysis model of a cooling loop. The reduction of the leak area amounts in the examined transient to ca. 25 % after 1 h and leads also to an approximately 23 % smaller leak rate. These results depend on the assumed position of the leak and on the initial leak size. Therefore, the results of accident analyses due to postulated leaks can be influenced by the change of the leak size during the course of the accident.
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