Development of a Typical Guidance to r-θ-z Modeling for 3-Dimensional Core Thermal-Hydraulic Analysis

As a computer performance has been remarkably improved and new numerical methods continuously developed, thermal-hydraulic assumptions for accident analysis, which were caused by deficient computer technology, running-time limitations and system complications, are not inevitable any more, and thus have been replaced with realistic modeling. The most unrealistic simplification in depicting nuclear reactor components has been arisen in a reactor core consisting of fuel assemblies, in-core detector guides, control rod drive mechanisms, core support materials, etc. Such are RELAP5 and most of the design codes for safety analysis that were developed in the past, which simplify the reactor core coolant to 1-dimensional flow channel. Safety analysis codes developed recently are able to describe a reactor core multi-dimensionally. For example, TRACE and MARS developed for regulatory audits are a 3-dimensional code. The 3-dimensional code basically considers r, θ and z flows. These treatments have difficulty that their numerical flow behaviors are highly dependent on a core modeling technique of a user, with a difference from a 1-dimensional code. The objective of this study is to propose a typical method for modeling a reactor core in the 3-dimensional accident analysis, in view of the fact that the 3- dimensional core modeling largely relies on a modeling skill of the users. This study, of which core modeling is based on best-estimated analysis, finally aims to produce its guidance