Simulation of CCFL in a PWR hot-leg pipe facility with CFD-3D

JOHAN SARACHE PIÑA; SANTIAGO F. CORZO; DARIO M. GODINO; DAMIAN E. RAMAJO

Belo Horizonte - Brasil

Congreso; Semana Nacional de Enngenharia Nuclear e da Energia e Ciências das Radiações - VI SENCIR; 2022

Universidade Federal de Minas Gerais e Departamento de Engenheria Nuclear

In the design and operation of nuclear plants, it is important to consider possible scenarios in the event of cooling circuit depressurization by loss of coolant accident (LOCA) of the reactor. Due to the high pressures in the coolant circuit, a break in any part of it could be crucial, causing the fast evaporation of the water in the core and the consequent dry-out of the fuel. In this scenario,the core has to be immediately reflooded by the Emergency Core Cooling System through the injection of cold water in the coolant circuit. But, it is expected that, during core flooding, a counter current steam-water flow be established, with water trying to flow toward the core and high-velocity steam flow leaving it. At this point the water circulation could stop because of the steam flow. This is known as the Onset of Counter Current Flow Limit (CCFL). In CCFL different two-phase flow regimes are present, but the segregated flow is predominant. Many researchers have tried to predict the state in which the Onset of CCFL occurs both by experimental test facilities, and also by numerical tools such as RELAP5 and detailed CFD-3D, mostly based on the two-fluid Eulerian method (TF). However, in a previous work we have demonstrated that the classic Volume of Fluid (VOF) method and sophisticated high order Piecewise Linear Interface Calculation (VOF-PLIC) techniques are significantly more accurate than the TF method to capture the free surface in complex disperse and segregated water-air problems. This work addresses the CCFL phenomenon with CFD-3D simulation in openFOAM using the VOF method to reproduce a well known CCFL test reported in literature. The comparison of numerical and experimental results allow us to conclude about the capability of the numerical models to predict the CCFL and the flow inversion before applying these models to solve a full scale nuclear installation