Development and verification of a neutronics-thermal hydraulics coupling code with unstructured meshes neutron transport model

Abstract

BackgroundThere is usually a strong coupling of neutronics-thermal hydraulics (N-TH) fields inside nuclear reactors.PurposeThis study aims to accurately simulate the multi-physics fields in nuclear reactors by developing a three-dimensional N-TH coupling code MORPHY tailored to advanced complex reactors.MethodsFirst of all, a three-dimensional triangular-z nodal variational nodal method (VNM) was employed for neutronics calculation. and the stiffness confinement method (SCM) was used to solve the neutron temporal-spatial equation; thermal hydraulic calculations were based on the one-dimensional multi-channel model and the one-dimensional cylindrical thermal conductivity model. Then, the accuracy of neutron dynamics was verified by TWIGL benchmark, Dodds benchmark, and the typical pressurized water reactor (PWR) benchmark NEACRP. Finally, the effects of different coupling methods and angle discrete orders on the results were analyzed and compared against reference solutions by PARCS.ResultsVerification results of TWIGL benchmark show that the deviation of relative power from the reference results is less than 0.5%. Compared with the results of Dodds benchmark, it verifies the MORPHY code's ability to describe unstructured meshes. The transient coupling calculation capability of MORPHY is verified by NEACRP benchmark.ConclusionsNumerical solutions by MORPHY are in good agreement with reference results of the TWIGL, Dodds and NEACRP benchmark problems. It is concluded that MORPHY can adapt to the transient N-TH coupling analysis of nuclear reactor cores.

Keywords

• neutron dynamics
• stiffness confinement method
• neutronics-thermal hydraulics coupling calculation
• transient
• transport code