ANALYSES OF SINGLE-PHASE HEAT TRANSFER AND ONSET OF NUCLEATE BOILING IN ROD BUNDLES
Thermal-Hydraulic (T/H) core code prediction of the existence and localization of boiling zones is crucial in the framework of axial offset anomaly (AOA) risk assessment of PWR cores. In this prospect, an experimental program -NESTOR-has been completed by Commissariat a lEnergie Atomique (CEA, France), Electricite de France (EDF, France) and Electric Power Research Institute (EPRI, USA). The aim of the NESTOR program has been to develop an accurate prediction model for the onset of nucleate boiling (ONB) boundary in a nuclear fuel bundle based on an ONB wall superheat criterion associated with a dedicated single-phase heat transfer model. The experimental scope of NESTOR program involved using two loops to measure axial velocities in sub-channels and heater rod surface temperatures, respectively, in identical 5x5 rod bundles. The first set of experimental measurements were devoted to a bundle configuration containing only simple support grids (SSG) in order to resemble, as closely as possible, a bare rod bundle. Test data analyses in this configuration have since been jointly carried out by the three NESTOR partners, each using its own T/H core code (FLICA IV for CEA, THYC-COEUR for EDF and VIPRE-I for US Penn State University on behalf of EPRI). This paper describes the analyses results and conclusions based on SSG configuration data. The data analyses methodology consisted of three successive stages: (I) T/H core code calibration -determination of specific input data related to the bundle configuration and required by further core code simulations of the tests; (Ⅱ) Single-phase heat transfer analysis -development of dedicated single-phase heat transfer models using single-phase test data along with sub-channel-averaged temperatures and velocities obtained from T/H code simulations. The heat transfer models were unique for each of the three codes and included both a heat transfer correlation and grid enhancement correction factor; (iii) ONB test analysis -assessment of an ONB wall superheat criterion based on stage (Ⅱ) models and, if necessary, development of a new ONB wall superheat criterion. The analyses showed that while the heat transfer models could correctly represent the single-phase test data, the ONB wall superheat is over-predicted by 1-3.5 K when compared to experimental values and open literature wall superheat correlations. However, the actual impact of this inconsistency on prediction of ONB boundary localization in this experimental SSG bundle configuration is low. Similar concurrent data analyses for tests on a mixing vane grid bundle configuration are under progress and results should be available by the end of 2010. NOMENCLATURE 1D, 2D, 3D Number of Dimensions A,B Coefficients for Friction and Grid Loss Coefficient Correlations AOA Axial Offset Anomaly EOHL End of Heated Length HTC Heat Transfer Coefficient LDV Laser Doppler Velocimetry MVG Mixing Vane Grids ONB Onset of Nucleate Boiling PWR Pressurized Water Reactor SSG Simple Support Grid T/H Thermal Hydraulic Dc Discrepancy Quantity used for Turbulent Mixing Model Optimization △P Pressure Drop.wo Darcy Friction Factor g(zg) Grid HTC Enhancement Correlation HDB Quadrant-Averaged Dittus Boelter Heat Transfer Coefficient Hexp Quadrant -Averaged Experimental Heat Transfer Coefficient ki Grid Loss Coefficient Nu Nusselt Number fwo Local Rod Heat Flux Pr Prandtl Number RDB Ratio of Experimental-to-Dittus-Boelter Heat Transfer Coefficients Re Reynolds Number p Density Tc Local Computed Sub-Channel Outlet Fluid Temperature Tf Local Computed Sub-Channel Fluid Temperature Tin Average Bundle Inlet Temperature Tm Local Measured Sub-Channel Outlet Temperature Tout Average Bundle Outlet Temperature Tsat,ONB Local Saturation Temperature at ONB Location Two Local Rod Surface Temperature Two,ONB Local Rod Surface Temperature at ONB Location V Axial Velocity vc Local Computed Fluid Velocity vm Local Measured Fluid Velocity vref Bundle Average Velocity Xc Dimensionless Computed Quantity Xm Dimensionless Measured Quantity Zg Distance to Upstream Grid
P.Péturaud R.K.Salko A.Bergeron M.N.Avramova
Electricité de France Chatou Cedex, France The Pennsylvania State University University Park, PA USA Commissariat à l’Energie Atomique Saclay, France
国际会议
18th International Conference on Nuclear Engineering(第18届国际核能工程大会 ICONE 18)
西安
英文
1-12
2010-05-17（万方平台首次上网日期，不代表论文的发表时间）