Steam Generator (SG) is one of the main components of the power cycle in pressurized water reactor (PWR), and it is the hub of primary coolant circuit and secondary circuit, so the thermal hydraulic analysis of the SG is crucial in the system design and safety analysis of the PWR. The horizontal steam generator (HSG) is one of the main types SG in the PWR nuclear power plant (NPP), and its advantages are that it has more secondary side water capacity and good safety and reliability. The VVER-1000 is a PWR with a thermal power of 3000 MW, and has four HSGs for four loops. The RELAP5 has been used to model the VVER-1000’s HSG and performs the analysis described in this paper. The HSG tube bundle is modeled by three horizontal channels, and the steam control volumes above the heat transfer tube bundle are modeled with three volumes. The steam space is modeled as a steam separator and the steam reception shield is the dryer. The HSG secondary side downcomers are represented with a separate component to provide the power of the natural circulation. To verify the accuracy of the model, three different typical conditions are simulated. The simulation results show that the model built in this paper can correctly simulate the operation of the HSG in VVER-1000.

The Brayton cycle using helium gas as a working medium has the advantages of high conversion efficiency, simple design, compact structure and high inherent safety. It can be combined with fossil energy, nuclear energy, solar energy and other forms of heat sources, and has broad prospects for development. The Helium Brayton cycle (HBC) is a highly integrated system. The dynamic characteristics of the Brayton cycle system are complex and difficult to obtain. It is necessary to establish the mechanism model of the HBC for research and analysis, and provide the basis for subsequent control system design. Based on the equations of conservation of mass, energy, momentum and energy balance, the thermo-hydraulic models of heat accumulator, precooler, intercooler and pipeline are established in this paper. The model was established according to the characteristic curves of helium turbine, low-pressure compressor and high-pressure compressor, and the shaft speed model was established according to the energy balance equation. Coupled with the above models, a lumped parameter nonlinear model of the HBC is established. Based on the MATLAB/Simulink platform, typical transient conditions such as external heat source power disturbance and external load disturbance are calculated. The results show that with the decrease of heat source input power, the temperature of each equipment and the net efficiency of the circulation decrease. When the external load fluctuates, the imbalance of compressor and turbine torque will lead to the rapid change of shaft speed.

Small pressurized water reactors (PWRs) have become a new trend in the current nuclear energy development due to their many advantages, such as compact equipment layout, high thermal efficiency, and strong cycle capability. Compared with large PWRs, small PWRs are designed to reduce the coolant inventory and increase the core power density, which is not good for nuclear safety. Severe accident studies on large PWRs cannot be directly applied to small PWRs. Loss of coolant accident is one of the main inducements of reactor core melting, which needs to be focused on prevention and treatment. It is of great significance for the safe operation of small PWRs to analyze and study severe accident induced by loss of coolant accident. In this paper, MELCOR is used to establish the severe accident analysis model of the primary loop system of a small PWR, and the loss of coolant accident is introduced to obtain the accident sequence from the shutdown of the reactor until the core degradation. At the same time, the core pressure, core liquid level and other key parameters are analyzed. The results show that in the case of a severe accident, compared with the large PWR, the small PWR takes a faster time for the pressure of the primary circuit and the containment pressure to reach equilibrium after the break accident occurs. The unbalanced radial power distribution causes the cladding of the 3rd ring to fail first. In the later stage of the severe accident, the melt plays a major role in heating the coolant. During the entire core degradation process, the upper fuel assemblies start to melt first, and the core does not completely collapse. The research results can provide reference for the formulation of severe accident management guidelines for small PWRs.