抄録
Three-dimensional coupled neutronics thermal-hydraulics reactor analysis is time consuming and occupies huge memory. A one-dimensional model is preferable than the three one in nuclear system analysis, control system design and load following. In this paper, a corewide three dimensional to one dimensional equivalent method has been developed. On the basis of this method 1D axial few groups constants were obtained. The equivalent cross sections were calculated by general spatial homogenization while the transverse buckling was computed through an equivalence based on the 3D flux conservation. Three steady test cases were performed on one dimensional finite difference code ODTAC and the results were compared with TRIVAC-5. The comparison shows that the one dimensional axial power distribution computed by ODTAC correlates well with the three dimensional results calculated by TRIVAC-5. In this study, DRAGON-4 was used to generate the few-group constants of fuel assemblies and the reflector few-group parameters were calculated by WIMS-D4. These collapsed few-group constants were tabulated in a database sorted in ascending order of fuel temperature, coolant temperature and concentration of boric acid. Trilinear interpolation was adopted in cross sections feedback during the transient analysis. In this paper, G1 rod drop accident(RDA) and G1 rod ejection accident(REA) were performed on ODTAC and the computation results were consistent of the physical rules.
