Development and validation of the code COUPLE3.0 for the coupled analysis of neutron transport and burnup in ADS

The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of(1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history,(2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and(3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces thevalidation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover,the proton irradiation module was verified by a comparison with the analytic method of calculating the210 Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.

Analysis of the axial fitting clearance between the fuel rod and the end seat in CIADS

The fuel assembly is key structure in China Initiative Accelerator Driven System,and the axial fitting clearance(AFC) for the fuel assembly design is an essential subject of study.In this paper,different methods are used to calculate critical stress in cylindrical shells.Because the thermal expansion of fuel assembly outer tube is larger than that of the cladding of fuel rod,enough space should be reserved between the upper end plug and upper seat slot.The collapse critical compressive stress of the cladding is obtained numerically through ANSYS simulation calculation.The AFC range between the fuel rod cladding and the end seat due to the displacement of thermal expansion is given by the theoretical formulas and ANSYS buckling analysis.These provide a reference for the AFC design of the reactor fuel assembly.