Development of fuel rod behavior analysis code and its application to supercritical CO2 cooled nuclear reactor.

• 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]