Numerical investigation of thermal–hydraulic characteristics in a steam generator using a coupled primary and secondary side heat transfer model

Abstract A coupled primary and secondary side heat transfer and thermal phase change model is used to investigate the thermal–hydraulic characteristics of a steam generator (SG) at Daya Bay Nuclear Power Plant (DBNPP). The simulation results reasonably reveal the ununiform boiling behavior in the secondary hot and cold legs. Vapor velocity is slightly higher than that of water and the corresponding slip-ratio first increases rapidly before gradually decreasing in the secondary hot leg and cold leg regions, a result which agrees with predictions using the drift-flux model. Cross-flow energy, which accounts for flow-induced vibration (FIV) at the U-bend tubes, is determined with the aid of localized thermal–hydraulic distributions, and the resulting FIV damage is predicted to be most severe at 0.35 m on the cold leg side and −0.2 m on the hot leg side of the U-bend region, respectively. These FIV damage predictions agree with measured plant data for the prototypical SG, showing that this model can provide the use information to improve thermal–hydraulic characteristics and help alleviate FIV damage in a SG.

[1]  B. Villard,et al.  Overview of Numerical Methods for Predicting Flow-Induced Vibration , 1988 .

[2]  S. Guentay,et al.  Potential steam generator tube rupture in the presence of severe accident thermal challenge and tube flaws due to foreign object wear , 2009 .

[3]  Guangxu Cheng,et al.  Determination of the critical crack length for steam generator tubing based on fracture-mechanics-based method , 2011 .

[4]  Jong Chull Jo,et al.  Fluidelastic instability analysis of operating nuclear steam generator U-tubes , 1999 .

[5]  N. Zuber,et al.  Average volumetric concentration in two-phase flow systems , 1965 .

[6]  Myung Jo Jhung,et al.  Dynamic characteristics of steam generator U-tubes with defect , 2005 .

[7]  J. Eoh,et al.  Development of a numerical analysis methodology for the multi-dimensional and multi-phase phenomena of a sodium–water reaction in an SFR steam generator , 2007 .

[8]  Stephen W. Webb,et al.  Simulation of the fluid–structure-interaction of steam generator tubes and bluff bodies , 2008 .

[9]  Heung June Chung,et al.  Flow-induced vibration of nuclear steam generator U-tubes in two-phase flow , 2011 .

[10]  Yuh-Ming Ferng,et al.  Thermal-Hydraulic Simulation of Localized Flow Characteristics in a Steam Generator , 2001 .

[11]  Yuanlong Yang,et al.  Numerically investigating the influence of tube support plates on thermal-hydraulic characteristics in a steam generator , 2013 .

[12]  A. Kebriaee,et al.  Numerical model for estimation of corrosion location in nuclear power plant steam generators , 2011 .

[13]  Zhang Yu.  Numerical Simulation on Vapor-Liquid Two-Phase Flow of the Secondary Circuit in Steam Generator , 2012 .

[14]  Zhi Zhang,et al.  Performance assessment for the water level control system in steam generator of the nuclear power plant , 2011, Proceedings of the 30th Chinese Control Conference.

[15]  Yuh-Ming Ferng,et al.  CFD investigating the impacts of changing operating conditions on the thermal-hydraulic characteristics in a steam generator , 2008 .

[16]  Alois Hoeld.  ICONE15-10206 APPLICATION OF THE MIXTURE-FLUID CHANNEL ELEMENT CCM WITHIN THE U-TUBE STEAM GENERATOR CODE UTSG-3 FOR TRANSIENT THERMAL-HYDRAULIC CALCULATIONS , 2007 .