The precise simulation of the reactor core either in steady state or during transients is very important for the safety assessment of nuclear power plants. This requires accurate determination of the parameters that influence the reactor operation. Coupling neutronic and thermal hydraulic schemes are developed to calculate these parameters. In the present paper, a coupling scheme between MCNP6 and ANSYS-FLUENT17.2 codes is proposed to obtain accurate radial and axial temperature distribution and hence pin power distribution for VVER-1000 fuel assembly. The Performance of the developed coupling scheme is investigated in steady state calculations. An iterative process is associated with the exchange of data between codes to meet the convergence criteria. The results obtained demonstrate that the proposed coupling scheme is able to simulate the VVER-1000 fuel assembly accurately. It gives information about thermal and neutronic behavior of the assembly and allows the feedback effects to be accurately modeled. This work is a step forward to establish a consistent methodology to be used in transient calculations.
Author contributions: All the authors have accepted responsibility for the entire content of this submitted manuscript and approved submission.
Research funding: None declared.
Conflict of interest statement: The authors declare no conflicts of interest regarding this article.
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