Genglei Xia

Investigation on Coupling Behaviors of Thermal-Hydraulics/Neutronics Under Asymmetrical Inlet Conditions

The reactor core is a complex system involving the reactor physics, thermal hydraulics and many other aspects. That means the distribution of the core power largely determines the profile of the thermal parameters, meanwhile the local thermal-hydraulics condition will in turn affect the neutronics calculation by moderator temperature effect and Doppler effects. Issues coupling the thermal-hydraulics with neutronics of nuclear plants still challenge the design, safety and the operation of LWR few years ago. Fortunately, the recent availability of powerful computer and computational techniques has enlarged the capabilities of making more realistic simulations of complex phenomena in NPPs.The current study deals with the development of an integrated thermal-hydraulics/neutronics model for Qinshan phase II NPP project reactor for the analysis of specific plant transients in which the neutronic response of the core is important, application of RELAP5-HD making use of the Helios code to derive the macroscopic cross-sections. Based on the coupled model, the steady state calculation and the transient simulation, involving the abnormal operation mode with asymmetrical coolant flux and temperature on the inlet of reactor, have been performed. The results show that the values obtained from coupled code RELAP5-HD calculation are in good agreement with the available experimental data, and the calculated accident parameters curves can predict all major trends of the transient. Steady state and transient condition calculation results are in accordance with the theoretical analysis from the aspect of coupled thermal-hydraulics/neutronics, this demonstrated a successful best estimate coupled RELAP5-HD model of Qinshan phase II NPP reactor has been developed, and the established model will provide a good foundation for the further analysis of the primary loop. It also can be concluded that the more accurate CFD method coupling three dimensional neutron kinetics code based on neutron diffusion method are necessary for steady-state calculation and analysis of transient/accident conditions when asymmetrical processes take place in the core. It is worth mentioning that RELAP5-HD code has already programmed the human-machine interface and the interface for coupling with other code, hence RELAP5-HD code has a broad application prospect in PWRs safety analysis.Copyright

Research on Flow Characteristics of Once-Through Steam Generator in Rolling Motion

Owing to its compact structure, the once-through steam generator (OTSG) is used widely in integrated Pressurized Water Reactor (PWR) The casing OTSG in concentric annuli tube is a new type of steam generator which applies double sides to transfer heat. The water of the secondary side goes through complicated phase change processes, and the flow pattern and heat transfer cases are much more complex than those of the natural circulation steam generator used in PWR. It is necessary to study their steady-state and dynamic characteristics. By means of THEATRe code which was based on the two-phase drift flux model and was modified by adding module calculating the effect of rolling motion, the casing OTSG simulation model in rolling motion was built. It can describe the parameters change in every section of OTSG accurately in rolling motion, and can embody dynamics characteristics from different aspects. By comparing the operational data in different rolling amplitude and rolling period, flow operational characteristics and principles were analyzed. The results can be used to analyze the thermal-hydraulics characteristics of the integrated PWR in rolling motion.Copyright

Analysis of Instability in Narrow Annular Multi-channel System Based on RELAP5 Code

The flow instability in narrow annular multi-channel system is analyzed in this paper using RELAP5/MOD3.4 code. The sensitivity of the number of the nodes and the number of the channels are studied firstly. It is found that the calculation results are sensitive to the number of control volumes and the number of the parallel channels. Base on optimized numbers, the two-phase flow instability between multi-channels (FIBM) is studied under different system pressures, inlet and outlet resistance coefficients, inlet sub-cooling, and the influence of them on the critical power are obtained.

Instability in Parallel Channel System

In this paper, the investigation on the instability in parallel channel system is summarized systematically. This phenomenon in parallel channel system is very typical, interesting and challengeable. The experiment data of a twin-channel system is used as the validation. Two typical methods are adopted to simulate this phenomenon for deciding the instability boundary. One is the integral method, which is based on the model of Clausse and Lahey and developed by Lee and Pan and GUO; the other is the classical system analysis code: Relap5/MOD3.4. In the experiment the influences of inlet resistance, system pressure and nonuniform heating are obtained. The influences of system pressure and inlet resistance can be simulated by both methods. However, there are some differences between the results of two methods. And for the effects of nonuniform heating and asymmetric inlet resistances, which are very popular in the nuclear power system, the results of numerical methods cannot get a good numerical agreement with those of experiment. It should be noticed in the practical engineering design. Finally, the typical “Ledinegg” instability phenomenon may occur in the parallel channel system according to the numerical results. Sometimes it will induce the burnout before the parallel channel instability. Both methods predict the same tendency. And a detailed explanation is given. The slope of the pressure drop-mass flux curve is the key to avoid the flow excursion phenomenon in parallel channel system.Copyright