Article ID | Journal | Published Year | Pages | File Type |
---|---|---|---|---|
8067215 | Annals of Nuclear Energy | 2018 | 12 Pages |
Abstract
This study describes a multi-scale nuclear safety analysis method for a main steam line break accident, which focuses on the 3-dimensional approach for the coolant temperature in the reactor pressure vessel related to the re-criticality and/or pressurizer thermal shock. The direct coupled code, CUPID/MARS, where CUPID is a 3-dimensional two-phase flow analysis code and MARS is a 1-dimensional nuclear system analysis code, was validated against a main steam line break test in the ATLAS facilities, which are a 1/2 scale integral test loop for a Korean APR1400 PWR. The calculation indicates that the suggested 4-step method, which expands the previous 2-step method of steady-state and transient calculations of a 1-dimensional system code, is valid for analyzing a direct 1-D/3-D coupled safety analysis, and the 3-dimensional temperature distribution in the downcomer agrees with the measured result.
Related Topics
Physical Sciences and Engineering
Energy
Energy Engineering and Power Technology
Authors
Ik Kyu Park, Seok Cho, Doo Hyuk Kang,