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For the floating pressurized water nuclear reactor, CHF is vital to the fulfilment of design and operation criteria for safe performance and the safety of nuclear fuel, and natural circulation is the primary circulation in many designs. However, the traditional method of look-up table or correlations evaluation is no longer applicable to the CHF characteristics because of additional force in the motion conditions, especially in natural circulation system (Gui et al., 2021). In this study, the prediction of CHF under motion conditions is carried out with code development and experimental study. The theory of natural circulation system models, subchannel models, critical heat flux models are coupled together with motion models to establish the prediction model for fuel rod critical heat flux characteristics in the natural circulation system under motion conditions. As rolling motion is studied in the previous study (Huang et al., 2018), inclination and heaving conditions are two motion conditions that are mainly considered in this study. The inclining and heaving conditions are modeled by adding the additional motion force. The code calculated results are validated with experimental data. The calculation results are analyzed with different inclining and heaving conditions, and the regularity of CHF variation characteristics is illustrated. • Models for CHF calculation in natural circulation system under motion conditions. • Development of a code for CHF calculation with the summarized models. • Validation of the code with the relative experimental study. • Analysis of the calculated CHF value under motion conditions in natural circulation. [ABSTRACT FROM AUTHOR]

• A more detailed list of all the models suitable for SCO 2 cooled nuclear reactor, and implant of all the models into FROBA. • Verification of the four models with other codes and experimental data. • The calculation results of the program FROBA- SCO 2 provide a reference for reactor core design. To simulate the thermal–mechanical performance of a Supercritical CO 2 -cooled reactor fuel during the long-term normal operation, material models and new thermal and mechanical models are implemented in FROBA code to develop FROBA-SCO 2. By calculation results comparison of the new models and ABAQUS, KAIST code and FRAPCON code, the reliability of new models were validated. The results show that with the effects of densification and burnup, the size and the temperature drop of the gas gap firstly increase and then decrease. The stress and the train of the cladding are affected by the pressure drop from both sides and change slightly. The fission gas release fraction is at a low level and the gas pressure increases slightly. The cladding corrosion behavior increases with the increment of temperature. The result indicated that the fuel element was safe enough during most time of the normal operation. The models established in this article calculates the fuel rod behaviors efficiently. [ABSTRACT FROM AUTHOR]