Assessment of TRACE CCFL Model with SBLOCA Experiment of IIST Facility

Jung-Hua Yang,Jong-Rong Wang,Hao-Tzu Lin,Chunkuan Shih

doi:10.1155/2012/731616

Abstract

In this paper, the TRACE model for IIST facility is developed and verified with the Small Break loss of coolant accident (SBLOCA) experiment of IIST (Institute of Nuclear Energy Research Integral System Test) facility. By using the Wallis and Kutateladze correlations of countercurrent flow limitation (CCFL) model, the TRACE analyses results, such as break flow rate, primary pressure, and the temperature of cold-leg and hot-leg, are consistent with the IIST data. The results show the Kutateladze correlation of CCFL model can well predict the SBLOCA behavior and present good agreement with IIST experiment data in this paper. Besides, the sensitivity study results of Kutateladze correlation in CCFL model are verified and compared with the IIST data.

Highlights

A reduced-height and reduced-pressure IIST facility has been established for safety studies of the Westinghouse three loops pressurized water reactor (PWR) since 1992
The countercurrent flow (CCF) may occur in the hot leg, the entrance to the steam generator, and the pressurizer surge line in IIST Small Break loss of coolant accident (SBLOCA) experiment
In order to verify the SBLOCA prediction of TRACE countercurrent flow limitation (CCFL) model, the comparisons of the results among Wallis correlation, Kutateladze correlation, and IIST data are considered in this paper

Summary

Introduction

A reduced-height and reduced-pressure IIST facility has been established for safety studies of the Westinghouse three loops pressurized water reactor (PWR) since 1992. When the CCFL occurs, the mass and heat transfer between gas and liquid phases reduces, and a water pool forms. This phenomenon prevents the core from cooling such that the fuel temperature rapidly increases [7]. One of the features of TRACE is its capability to model the reactor vessel with 3D geometry. It can support a more accurate and detailed safety analysis of nuclear power plants. The sensitivity studies of CCFL model in water level and core cladding temperature are discussed

IIST Facility and SBLOCA Experiment

CCFL Model in TRACE

Results and Discussions

Conclusions