Volume 84, Issue 3

Accurate prediction of the critical heat flux (CHF) is one of the key tasks of PWR core design and safety assessment, for the maximal allowable heat flux in the reactor core is limited by CHF. Since CHF in rod bundle cannot be predicted analytically, up-to-date predictive approach is based on empirical correlations related to the local thermal-hydraulic conditions, geometry and power distribution. However, development of CHF correlation for PWR fuel assemblies under low pressure conditions (2–10 MPa) is constrained by limited amount of experimental data points, which builds up in statistics a typical problem of small sample amounts, but requiring simultaneously high prediction accuracy. In our previous study, stepwise regression method was applied to develop a dimensional, empirical CHF correlation for PWR under low pressure conditions, termed as the advanced low pressure CHF correlation (ALPC), which successfully solves the challenge of small sample problem. However, the ALPC correlation still uses dimensional independent variables with less physical meanings, which limits its physical interpretability. In the current study, stepwise regression method was used to develop a revised, dimensionless version of the ALPC CHF correlation. First, various dimensionless, two-phase thermal-hydraulic parameters that might influence CHF were selected as candidate independent variables. With stepwise regression, the form and coefficients of the revised CHF correlation were optimized in a dynamic manner. Compared to the current ALPC correlation, the revised version developed in this study possesses a similar simple form but a much higher prediction accuracy. Revision of the ALPC correlation demonstrates clearly the advantages of utilizing dimensionless parameters as independent variables in CHF correlation, which points out a new direction of developing rod-bundle CHF correlations for engineering purpose.

Experimental investigation on adiabatic transient flow fluctuation characteristics in a 3 × 3 rod bundles subjected to rolling motion in normal pressure and temperature was carried out. Single-phase flow as well as gas parameters under stagnant liquid condition were concerned. Ranges of liquid Reynolds number and gas flow rate were 657–3 328 and 0.079–0.2 m 3 /h, respectively. Rolling period and amplitude range from 8 s to 16 s and from 5° to 15°, respectively. Fluctuation amplitudes of transient parameters such as flow rate, pressure drop, void fraction and gas rising velocity were focused on this study. The results indicate that flow rate and frictional pressure drop present obvious fluctuations after rolling motion, and the fluctuation amplitudes decrease with the decrease of bypass valve opening. Flow rate reaches to peak at the positive rolling amplitude and it reaches to trough at the negative rolling amplitude, and there is no obvious phase difference between the fluctuation and rolling angle. Periodical fluctuation of gas flow rate was weakened with the increase of gas inlet pressure and valve opening under experimental conditions, while the variations of void fraction and gas rising velocity were limited. Transient parameter fluctuations correspond to the rolling period at low gas flow rate, whereas their periodical fluctuations are inconspicuous in relative high gas flow rate. Amplitudes of void fraction and gas rising velocity increase with increasing the rolling amplitude while the amplitudes of void fraction and gas rising velocity are nearly invariable with the rolling period no matter if gas flow rate oscillates.

Kuosheng Nuclear Power Plant (NPP) is a BWR/6 plant in Taiwan. First, this study focuses on the establishment of the TRACE/SNAP model for Kuosheng. To check the system response of the Kuosheng TRACE/SNAP model, the Final Safety Analysis Report (FSAR) and startup tests data are used to assess the TRACE/SNAP model. The TRACE predictions are consistent with the FSAR and startup tests data. This indicates that there is a respectable accuracy in the TRACE/SNAP model of Kuosheng NPP. Second, this study also focuses on the application of the TRACE/SNAP model in the core shroud leakage. The Kuosheng NPP TRACE/SNAP model is used to perform the analysis of core shroud leakage. The TRACE results imply that Kuosheng NPP is in a safe situation when the core shroud leakage transient occurs.

In this paper, development of coupled codes using two-group neutron diffusion kinetics code and computational fluid dynamics (CFD) solver Fluent has been introduced. Way of coupling, time step control algorithm and spatial mesh overlays have been summarized in detail which are basic components and challenges of the coupling methodologies. The implement and verification of coupled code have been modeled on integral pressurized water reactor (IPWR) IP200 with hexagonal fuel assembly in the core and once-through steam generators. The steam line break core transient was analyzed in coupled code simulation of a core boundary conditions derived from system code simulation results. The results presented transient three-dimensional distribution of the key operation parameters such as reactor power and coolant temperature, also demonstrated the inherent safety features of IP200. The current work will bring about the ability to explore multi-scale and multi-dimensional safety transient evaluations and give more precise neutronics/thermal-hydraulics mapping.

Numerical simulations of selected subchannel tracer (Potassium Nitrate) based experiments have been performed to study the capabilities of state-of-the-art of Computational Fluid Dynamics (CFD) codes. A Reynolds stress model (RSM) has been selected as the primary turbulence model to be applied for the simulation case as it has been previously found reasonably accurate to predict flows inside rod bundles. As a comparison, the case is also simulated using a standard k-∊ turbulence model that is widely used in industry. Flow enters into the computational domain through the mass inflow at the three subchannel faces. Simulation results have been extracted at different locations of the mixing zone and downstream zone. The local mass fraction shows uniform mixing. In present investigation using the species transport model, includes the study of diffusion of two fluids (water and KNO 3 ) to study its interaction and how mixing is enhanced. The species model is concerned with the transport of species with in this phase, where it is experimentally found that our species are completely miscible. The effect of the applied turbulence model is nearly negligible just before the outlet plane because the distributions look like almost identical and the flow is fully developed. On the other hand, quantitatively the dimensionless mixing scalar distributions change noticeably, which is visible in the different scale of the colour bars.

Control rods provide adjustable manipulation of reactivity and prevent fission reaction from accelerating beyond control. In this paper, the Monte Carlo Code, MCNP6 was used to study the effect of using various control rod materials on the main safety parameters of a typical plate type material test research reactor. Three types of control rod materials were considered, Ag-In-Cd, B 4 C, and Hafnium. Parameters investigated are reactivity worth, excess reactivity and shutdown margin of absorber materials. The model was used to evaluate the effect of various materials on the flux and power shaping of the core. The resulting multiplication factor for this study was compared with previous work and found acceptable. It was found that boron carbide has the largest effect on safety parameters, as well as flux and power distribution.