Analysis of the NUPEC PSBT Tests with FLICA-OVAP

Matteo Bucci ,Philippe Fillion

doi:10.1155/2012/436142

Abstract

This paper discusses the results of a computational activity devoted to the prediction of two-phase flows in subchannels and in rod bundles. The capabilities of the FLICA-OVAP code have been tested against an extensive experimental database made available by the Japanese Nuclear Power Energy Corporation (NUPEC) in the frame of the PWR subchannel and bundle tests (PSBT) international benchmark promoted by OECD and NRC. The experimental tests herein addressed involve void fraction distributions and boiling crisis phenomena in rod bundles with uniform and nonuniform heat flux conditions. Both steady-state and transient scenarios have been addressed, including power increase, flow reduction, temperature increase, and depressurization, representative of PWR thermal-hydraulics conditions. After a brief description of the main features of FLICA-OVAP, the relevant physical models available within the code are detailed. Results obtained in the different tests included in the PSBT void distribution and DNB benchmarks are therefore reported. The relevant role of selected physical models is discussed.

Highlights

Based on Nuclear Power Energy Corporation (NUPEC) PWR subchannel and bundle tests (PSBT), an international benchmark has been promoted by OECD and NRC and has been coordinated by Penn State University [1]
This paper has presented the current capabilities of the FLICA-OVAP code in predicting void distribution and boiling crisis phenomena
The NUPEC database released in the frame of the OECD/PSBT international benchmark has been addressed

Summary

Introduction

Based on NUPEC PWR subchannel and bundle tests (PSBT), an international benchmark has been promoted by OECD and NRC and has been coordinated by Penn State University [1]. The aim of this benchmark is to encourage advancement and assessment of numerical models in subchannel analysis of fluid flow in rod bundles, which has very important relevance for the nuclear reactor safety margin evaluation. To provide a relevant answer to different core concepts and multiple industrial applications, several models coexist in the FLICA-OVAP platform: the homogeneous equilibrium model, the four equations drift flux model, the two-fluid model, and a general multifield model, with a variable number of fields for both vapor and liquid phases. An adapted set of closure laws is proposed concerning heat and mass transfer, interfacial and wall forces, and turbulence

The FLICA-OVAP Code

Prediction of the Boiling Crisis

Void Distribution Results

DNB Results

Steady-State Series

Conclusions and Perspectives