Coupling of Neutronic and Thermal-hydraulic Calculations for Tehran Research Reactor Core Analysis

摘要

Tehran Research Reactor (TRR) is a multipurpose research reactor with 5 MW power level, that is usedfor a wide variety of research activities. It is of high interest to have a set of reliable coupled neutronic and thermal-hydraulic calculation tools that could be used for prediction of core behavior in different core applications. For neutronic part, two different calculation lines were used to simulate the core behavior and perform the necessary neutronic calculations. First combination of cell calculation transport code WIMS-D/4 and three dimensional core calculation diffusion code CITVAP were used to and next a Monte Carlo code MCNP-4B were used to cross check the results. For thermal-hydraulic part, the calculations have been done with CAUDVAP and TERMIC computer codes. The impact of the statistical treatment of uncertainty factors on the main thermal-hydraulic parameters has also been considered here. In order to assess the extent of the practical feasibility of our calculation procedures and models and the reliability of our calculation tools, the core analysis has been applied for the First Operating Core (FOC) of TRR and also for some subsequent cores for which a set of full comprehensive data exist and checked against them. The comparison shows a rather good agreement between them. After making sure of the adequacy of the models and the tools, the coupled neutronic and thermal-hydraulic codes have been successfully applied for other aspects of core analysis.