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A versatile reactor system analysis code RESYS for advanced nuclear reactor was developed by using object-oriented language C++. Based on RESYS code, the model of thermionic nuclear reactor TOPAZ-Ⅱ developed in the former Soviet Union was established, and its start-up process was simulated. The established TOPAZ-Ⅱ reactor system model included reactor core thermal model, thermionic conversion system model and the radiator model. The current density model of cesium thermionic converter used model by Rasor. And the 6 groups of delayed neutron point reactor kinetic model was used to consider the influence of various structural components on reactivity. The point reactor equation with strong rigidity was solved using the Gear algorithm in the RESYS code. The reactivity feedback model of TOPAZ-Ⅱ reactor took into account the Doppler reactivity feedback of the fuel, the reactivity feedback of the electrodes, the reactivity feedback of the moderator and reflector, and the introduction of reactivity control drums. The steady state and start-up transient were simulated based on established thermal-hydraulic model and RESYS code. At steady state, the maximum temperature on the inner surface of the central thermionic fuel element is 2 291 K, the maximum temperature on the outer surface is 2 066.55 K, the maximum temperature on the emitter is 1 961 K, the maximum temperature difference between the electrodes is close to 1 200 K, the temperature at the coolant inlet is 743 K, and the temperature at the core outlet is 837 K. The comparison between the steady-state operating conditions calculated by the RESYS code and the TOPAZ-Ⅱ design values was carried out. The calculated steady-state electric power output is consistent with the calculation results of TITAM code. The calculation results verify the correctness of the developed RESYS code and the established TOPAZ-Ⅱ system model. During the startup process, the reactivity feedback of the moderator and reflection layer is positive feedback, which dominates all feedback effects. The overall reactivity feedback of the TOPAZ-Ⅱ reactor core is positive and does not have a self-stable negative feedback effect, which increases the difficulty of reactor control. During the start-up process of TFE, due to changes in the gas thermal conductivity of the electrode gap, the temperature of the fuel pellets and emitter rapidly increases. At this time, some of the thermal energy generated by fission is stored in the fuel pellets, causing a decrease in coolant temperature and the power of the radiation radiator.

As the fourth generation advanced reactor, the Brayton cycle system of small helium-xenon cooled reactor has the advantages of light weight, compact structure, simple system, fast start or stop, and is very suitable for the energy supply of remote areas and islands. Nuclear power plant is radioactive and highly hazardous after accidents, so it is very important to study its safety characteristics. In addition, helium-xenon cooled reactor is an advanced future reactor type, and its accident safety characteristics are not yet clear. Therefore, it is necessary to carry out the safety characteristics research of Brayton cycle system of small helium-xenon cooled reactor, especially under the condition of unprotected control. Firstly, based on Modelica language, through the modular dynamic modeling of key equipment, the accident safety characteristics analysis program of helium-xenon cooled reactor Brayton cycle system was developed. Secondly, the steady-state rated condition of the system was simulated, and the reliability of the accident safety characteristics analysis program of the system was preliminarily verified by comparing the error between the simulation value and the design value of the steady-state rated condition of the system, in which the maximum relative error is 2.31%. Finally, the transient conditions of various system accidents were simulated, and the dynamic response characteristics of transient conditions such as step introduction positive reactivity accident, step introduction negative reactivity accident, coolant incorrect charging accident and coolant incorrect discharging accident were analyzed. The results show that the simulation results of the system analysis program are reasonable. The reactor system has a certain self-stability and self-adjustment ability for step introduction positive reactivity accident and coolant incorrect charging accident under the condition of constant load and unprotected control. Under the conditions of step introduction positive reactivity accident and coolant incorrect charging accident, after the rotational speed increases by 21.6% and 12% respectively, it can be stabilized in another steady-state without protection and control. However, under the conditions of constant load and unprotected control, the reactor system can not be stabilized under another steady-state condition through self-regulation for the step introduction negative reactivity accident and the coolant incorrect discharge accident. Under the conditions of step introduction negative reactivity accident and coolant incorrect discharge accident, the corresponding rotational speed drop rate at 800 s is 85 rpm/s and 10 rpm/s respectively. It is necessary to control the rotational speed drop rate of the rotor through some protection and control measures to ensure the safety of reactor system. The relevant results provide a theoretical basis for the safety analysis of Brayton cycle system of helium-xenon cooled reactor.

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.