Development of circuit model with natural circulation of coolant under conditions of ship motion

Safety is the key requirement to any nuclear power installation. Various factors affect safety during operation of the nuclear power installation. These factors are difficult to study due to the high economic costs. This problem can be solved by developing prototype models to conduct the research of many complex processes. Dynamic impact on the ship installation is one of these processes. The most significant impact is the impact on the natural circulation of the coolant, that is one of the basic emergency safety systems. Also, it is a promising way to ensure movement in the main circulation circuit. The purpose of this paper is to assess the influence of external dynamic forces on the processes of natural circulation. For the study a testing bench has been developed that simulates one of the circulation loops of the reactor unit. The basic method to obtain experimental data is temperature sounding of the specific sections of the circulation route. A mathematical model has been developed that describes this process. The model is based on the equations of momentum conservation and heat balance. In accordance with the experimental data, the calculation of natural circulation for static and dynamic modes has been carried out. A mathematical model to describe this process has been developed. A comparative analysis of the results of calculating the static and dynamic modes has been carried out. It is founded out that the decrease of mass flow rate is about 10 % as compared with the static regime. It confirms the qualitative effect of ship motion on natural circulation. The practical significance of the research is the development of a model under conditions of ship motion, as well as verification of the model at the testing bench. The results show a significant effect of ship motion on the mass flow rate of the coolant in the case of natural circulation. Thus, to ensure the required safety of ship installations, it is recommended to conduct a study of natural circulation in accordance with the developed model under conditions of maximum possible ship motion.

Possibility of simulating natural circulation in fast neutron reactors using a light water test facility

The paper evaluates the possibility of modeling the heat transfer phenomena in a liquid-metal coolant using a light water test facility. It considers the natural circulation of the coolant in the upper plenum of the fast-neutron reactor. The sodium-cooled BN-1200 reactor was selected as the reactor installation to be modeled. The development of novel reactor designs must be based on the results of experimental studies. Some problems of modeling thermohydraulic processes in BN type reactors are studied by using sodium test facilities. Experimental studies of natural convection processes using light water test facilities can be considered as a good alternative to those using sodium test facilities. To validate the model, the similarity theory and the “black box” method were used and their principles and applicability were analyzed. Using the “black box” method makes it possible to avoid detailed modeling of such components as the reactor core and heat exchangers, replacing them by a simplified representation of these components to simulate the integral characteristics of the existing real life equipment. The paper considers the basic criteria which determine the similarity of the thermohydraulic processes under study. The governing criteria of similarity were estimated based on the fundamental differential equations of natural convection heat transfer. Based on these criteria, a set of dimensionless values was obtained which show the correlation between the model parameters and the characteristics of the reactor facility. Besides, generalized relationships were derived which can be used to estimate the scaling factors for calculating the key values of the reactor facility based on the model parameters. These relationships depend on the thermal-physics parameters of the working fluids, the geometrical scale value and the ratio of the thermal power of the model to that of the reactor facility, i.e., model-to-reactor thermal power ratio. The conditions under which it is possible to model sodium coolant by light water with adequate accuracy were analyzed. An example is given of the numerical values of the scaling factors for one of the reference light water test facilities. The paper uses the experience of a number of foreign researchers in this field, in particular, the accepted assumptions which do not result in serious loss in modeling accuracy. According to the available estimates, the assumptions used do not result in considerable losses in accuracy. Thus, the natural circulation of the sodium coolant in the upper plenum of the fast-neutron reactor can be simulated with adequate accuracy by using light water test facilities.