The development of multi-physics approach with Monte Carlo and computational fluid dynamics coupling for reactor cores

Kutbay F., Şentürk Lüle S.

Nuclear Engineering and Design, vol.402, 2023 (SCI-Expanded) identifier identifier

  • Publication Type: Article / Article
  • Volume: 402
  • Publication Date: 2023
  • Doi Number: 10.1016/j.nucengdes.2022.112127
  • Journal Name: Nuclear Engineering and Design
  • Journal Indexes: Science Citation Index Expanded (SCI-EXPANDED), Scopus, Academic Search Premier, Aerospace Database, Applied Science & Technology Source, Chemical Abstracts Core, Communication Abstracts, Computer & Applied Sciences, INSPEC, Metadex, Pollution Abstracts, Civil Engineering Abstracts
  • Keywords: Multi-physics, Computational fluid dynamics, Monte Carlo method, Loose coupling, Cross section generation, Thermal scattering data generation
  • Istanbul Technical University Affiliated: Yes


© 2022 Elsevier B.V.Highly accurate solutions for neutronic and thermal hydraulic phenomena in reactor cores certainly improve the safety conclusions about the reactors. The coupling methodology of governing physics of these two fields by using Monte Carlo and computational fluid dynamics (CFD) methods was described in this study with its application to Istanbul Technical University (ITU) TRIGA Mark II Research Reactor. As a different approach, instead of not modelling top and bottom grid plates of the reactor that is the common practice for geometrical simplification, the effect of grid plates on flow hydrodynamics was included into the conjugate heat transfer calculations as boundary conditions. The investigation on the effect of core element temperature, cross sections, and S(α,β) thermal scattering data showed the need of proper coupling of neutronic and thermal hydraulic fields since 30–45% relative error on excess reactivity value of the ITU research reactor was observed. With the developed loose coupling model, 483 individual cross section and 483 different S(α,β) thermal scattering data were generated by using NJOY nuclear data generation code and MAKSXF and OTF utilization tools of MCNP6.2 code. The reactivity loss against power increase in research reactor value of this study showed 6% relative error with loose coupling comparing to maximum 50% without coupling.