Yingwei Wu

Sub-Channel Analysis of Pb-Bi-Cooled Reactor With Modified COBRA-EN

Lead-bismuth eutectic (LBE) cooled fast reactor, one of the six types of reactors in Gen-IV, has very good inherent safety and significant advantages in reducing and burning nuclear wastes, enhancing economy. Also LBE cooled accelerator driven system (ADS) has been a very innovative and potential waste burner. COBRA-EN is a mature, stable and widely-used sub-channel analysis code for light water cooled reactor but it couldn’t be applied in Pb-Bi-cooled reactor directly. Some modifications were made for COBRA-EN in the present work, then the code was named COBRA-PB and was suitable for the sub-channel analysis of Pb-Bi-cooled reactor. The modified code was verified and validated with CFX and experimental results. There was a good agreement between the two results. Then sub-channel analysis of Pb-Bi-cooled reactor was done with the modified code.Copyright

Critical Power and Void Fraction Prediction of Tight Bundle Designs

Abstract With the objective of providing long-term energy supply via actinide breeding and burning, the next-generation boiling water reactor (BWR) design, the Hitachi’s resource-renewable BWR (RBWR), has been proposed. Unlike a traditional square lattice BWR fuel bundle, the RBWR bundles are shorter with hexagonal tight lattice arrangement and heterogeneous axial fuel zoning. The RBWR’s different core geometry combined with the higher power-to-flow ratio and void fraction necessitates the reexamination of the standard BWR thermal-hydraulic models. For the prediction of dryout, the previously derived best-estimate empirical correlation showed significant scatter when compared to experimental data within its calibration database. In this work, the correlation is further calibrated and improved by supplementing tight bundle data with relevant critical power data for tubes and annuli to better quantify the effects of various parameters and by incorporating subchannel-level results to account for intra-assembly flow mixing. Another approach using the mechanistic three-field model is also investigated, and the minimum critical power ratio of the RBWR design is evaluated. For the prediction of void fraction, measurements and the three-field model in annular flow regime reveal that the common drift flux approaches tend to overestimate the void fraction at small hydraulic diameters. The void fraction dependence on hydraulic diameter below 10 mm requires further experimentation and high-fidelity mechanistic simulations.

Experimental Study on the Thermal Hydraulic Characteristics of Liquid Sodium Flowing in an Annulus

Thermal hydraulic characteristics of liquid sodium flowing in an annulus are experimentally studied. The annulus is 1100 mm in length, 6 mm as inside diameter and 10 mm as outside diameter. The heat flux in the experiment is from 50 to 210 kW/m2, with Re number from 0 to 18000 and average fluid temperature from 200 °C to 500 °C. Experimental data show that the flow regime of liquid sodium flowing in the annulus can be divided into three regions including laminar flow (Re 4000). The effects of heat flux, Re number and average fluid temperature of the test section on the heat transfer coefficient are investigated separately. For different regions, correlations for the friction coefficient and for the Nu number are obtained from the experimental data.Copyright

A Prediction of the Leakage Through Cracks for Leak Before Break

A code was developed in this study to predict the leakage of the leak before break (LBB). Various stagnation conditions were considered, including the subcooled water, the two-phase fluid and the overheated steam. Moreover, both the critical and noncritical flow was studied. The Henry-Fauske critical flow model was revised by a new phase transition point and the pressure drop due to friction and turns were modified. The code was verified by the comparison with the experimental data on the leakage of conventional pipes, artificial cracks and naturally occurring cracks, which shows a good agreement and this code has a higher precision than the existing codes. The influence of crack morphologies on LBB leakage was discussed, including the local roughness, the global roughness, the crack open displacement (COD) and the number of the corners. Besides, the dependence of the LBB leakage on stagnation enthalpy and back pressure was also investigated.Copyright

The Influence of Non-Uniform Heating on Two-Phase Flow Instability in Parallel Channels

The two-phase flow instability in parallel channels heated by uniform and non-uniform heat flux has been theoretically studied in this paper. Based on the homogeneous flow model in two-phase region, the system control equations of parallel channels were established. Semi-implicit finite-difference method and staggered mesh method were used to discretize the system control equations and the difference equations were solved with a chasing method. The cosine profile and uniform constant heat flux represent the non-uniform and uniform heating condition, respectively. The marginal stability boundaries (MSB) of parallel channels and the three-dimensional instability spaces (or instability reefs) of different heat flux models were obtained. For cosine profile heating, the stability of parallel channels increases with the increase of the system pressure and inlet resistant coefficient. In high inlet subcooling region, cosine heat flux can strengthen the system stability. However, in low inlet subcooling region, the negative effect to system stability will be caused by non-uniform heating. The increase of inlet resistant coefficient will move the turning point of the MSB to high inlet subcooling number.Copyright

Transient Study on Sodium Heat Pipe in Passive Heat Removal System of Molten Salt Reactor

High temperature heat pipes are effective devices for heat transfer, which are characterized by remarkable advantages in conductivity, isothermality and passivity. It is of significance to apply heat pipes on new concept passive residual heat removal system (PRHRS) of molten salt reactor (MSR). In this paper, the transient performance of high temperature sodium heat pipe is simulated with numerical method in the case of MSR accident. The model of the heat pipe is composed of three conjugate heat transfers, i.e. the vapor space, wick structure and wall. Based on finite element method, the governing equations and boundary conditions are solved by using FORTRAN code to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results indicated that high temperature sodium heat pipe had a good operating characteristic and removed the residual heat of fuel salt rapidly under the accident of MSR.Copyright