DIFFERENT SIMULATIONS OF THE PHASE 2 OF THE OECD/NRC BWR TURBINE TRIP BENCHMARK WITH THE CODE DYN3D

摘要

The phase 2 of the OECD/NRC BWR TT Benchmark was analysed with the core model DYN3D by using the specified core thermal-hydraulic boundary conditions. The results of the standard case that includes the consideration of one thermal-hydraulic channel per assembly, the ADF and the standard phase slip model of DYN3D show a good agreement with measured values as the radial averaged power distribution and the total power versus time. The influence of the ADF is small, if these core-averaged values are compared. Differences in the results of single fuel assemblies were observed. If the core is described by only 33 thermal-hydraulic channels, it is observed too. The ZUBER-FINDLAY slip model shows only small deviations to the standard model of MOLOCHNIKOV. The one-dimensional model also describes the average core values in good agreement with the standard model. If the transient is calculated by the coupled code system, the core behaviour is strongly influenced by the interaction of the thermal hydraulics of the core with the remainder of the system. Nevertheless the neutron kinetics and thermal hydraulics of the reactor core should be described as detailed as possible. This is important in the case that local and/or extreme values are considered. If a fast-running one-dimensional model is applied, the one-dimensional cross section sets have to be elaborated by radial condensing. However, due to the fast computers nowadays available the man power needed for generating the one-dimensional cross section library can be saved by applying three-dimensional core models.