论文信息 - Thermal hydraulic performance assessment of dual-cooled annular nuclear fuel for OPR-1000

Thermal hydraulic performance assessment of dual-cooled annular nuclear fuel for OPR-1000

Abstract An internally and externally cooled annular fuel was proposed for an advance PWR, which can endure substantial power uprating. KAERI is pursuing the development for a reloading of power uprated annular fuel for the operating PWR reactors of OPR-1000. In this paper, the characteristics and verification of the MATRA-AF are described. The thermal hydraulic performance of a 12 × 12 annular fuel is calculated for the major design parameters and its performance is compared against the reference 16 × 16 cylindrical fuel assembly. In particular, the enhancements of the thermal hydraulic performance of dual-cooled annular fuel are estimated for the 100% normal power reactor core. The purpose of this study is to estimate a normal power for OPR-1000 with dual-cooled annular fuel, and ultimately to assess the feasibility of 120% core power. The parametric study was carried out for the fuel rod dimension, gap conductance, thermal diffusion coefficients, and pressure loss of the spacer grids. As a result of the analysis on the nominal power, annular fuel showed a sufficient margin available on DNB and fuel pellet temperature relative to cylindrical fuel. The margin amount seems accommodating a 20% power-uprate seems viable.

[1] F. Dittus,et al. Heat transfer in automobile radiators of the tubular type , 1930 .

[2] F. de Crécy. The effect of grid assembly mixing vanes on critical heat flux values and azimuthal location in fuel assemblies , 1994 .

[3] Mujid S. Kazimi,et al. Introduction to the Annular Fuel Special Issue , 2007 .

[4] J. M. Healzer,et al. A void fraction correlation for generalized applications , 1992 .

[5] Soon Heung Chang,et al. Development of a thermal-hydraulic analysis code for annular fuel assemblies , 2003 .

[6] B. A. Zolotar,et al. Mechanistic model for predicting two-phase void fraction for water in vertical tubes, channels, and rod bundles. [PWR; BWR] , 1982 .

[7] Soon Heung Chang,et al. Development of a thermal hydraulic analysis code for gas-cooled reactors with annular fuels , 2006 .

[8] P. Richardson,et al. Boiling heat transfer and two-phase flow , 1956 .

[9] S. Levy,et al. Steam Slip—Theoretical Prediction From Momentum Model , 1960 .

[10] Tae Hyun Chun,et al. Evaluation of a Dual-Cooled Annular Fuel Heat Split and Temperature Distribution , 2009 .

[11] Mujid S. Kazimi,et al. Thermal-Hydraulic Design of High-Power-Density Annular Fuel in PWRs , 2007 .

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