DEVELOPMENT OF A METHODOLOGY FOR MODELING THE DYNAMIC VOLUMETRIC DISTRIBUTION OF FUEL IN AN EXPERIMENTAL DEVICE USING MCNP

The paper presents a methodology for modeling the dynamics of volumetric fuel distribution based on the coupling of thermal-hydraulic and neutronic calculations. An algorithm was developed for the automated conversion of CFD simulation results into an MCNP computational model. The methodology makes it possible to account for changes in geometry, phase composition, and spatial distribution of fuel and structural materials during fuel degradation and melting processes. To achieve this, software was developed for processing thermal-hydraulic calculation data, generating computational lattices of various types, implementing an angular duplication algorithm, and automatically creating input files for neutronic calculations. Unlike the conventional approach, which relies on the development of a set of separate neutronic models corresponding to fixed stages of fuel degradation, the proposed methodology enables power distribution calculations for any state of the experimental device obtained from thermal-hydraulic simulations. Verification of the developed neutronic model of the test facility containing fuel within the reactor model of the IGR reactor was performed. The obtained results can be used for computational support of experiments investigating nuclear fuel degradation and melting processes, as well as for the development of coupled thermal-hydraulic and neutronic modeling methodologies for reactor research applications.

1. ANSYS, Inc. Products Release 2021 R1. Academic research Mechanical and CFD.

2. Suraev A.S., Vityuk G.A., Pakhnits A.V., Irkimbekov R.A. Computer Modeling of Severe Accidents at the IGR Reactor. In: Proceedings of the 10th International Conference “Nuclear Energy, Nuclear and Radiation Technologies of the XXI Century”, Minsk, May 28-31, 2024. Minsk: Joint Institute for Power and Nuclear Research Sosny of the National Academy of Sciences of Belarus, 2024. P. 50.

3. Kelsingazina R., Vityuk V., Vurim A., Vityuk G., Mukhamedov N., Tikhomirov G. Computational approaches for determining the nuclear heating value of structural materials during the irradiation at the IGR reactor // Annals of Nuclear Energy.– 2024.– Vol.204, art. no. 110532.– https://doi.org/10.1016/j.anucene.2024.110532.

4. Ivanov K., Avramova M. Challenges in coupled thermal– hydraulics and neutronics simulations for LWR safety analysis // Annals of Nuclear Energy.– 2007.– Vol. 34, No. 6.– P. 501–513.– https://doi.org/10.1016/j.anucene.2007.02.016

5. Avramova M., Ivanov K. Coupled Multiphysics Modeling for Nuclear Reactor Applications // Nuclear Engineering and Design. – 2010.– Vol. 240.– P. 601-614. https://doi.org/10.1016/j.pnucene.2010.03.009

6. Ruslan A. Irkimbekov, Vladimir A. Vityuk, Galina A. Vityuk, Olzhas M. Zhanbolatov, Artur S. Surayev, Yuri A. Popov, Andrey N. Kotlyar (2025) Development of a new computational support tool for experiments in the impulse graphite reactor // Nuclear Engineering and Design, 443 (2025), 114278, doi.org/10.1016/j.nucengdes.2025.114278

7. MCNP-6.1 Monte-Carlo N-Particle Transport Code: Los Alamos National Laboratory.

8. Kabdylkakov E.A., Suraev A.S., Zhanbolatov O.M., Vityuk G.A., Irkimbekov R.A. Improvement of Methods for Modeling the Melting of Fuel Assemblies under Experimental Conditions at the IGR Reactor // NNC RK Bulletin.– 2025.– Issue 2(102).– P. 151–164.– https://doi.org/10.52676/1729-7885-2025-2-151-164.

9. Eickhoff M., Rückert A., Pfeifer H. Solidification modeling with User defined function in Ansys Fluent //Progress in Applied CFD – CFD2017 Selected papers from 12th International Conference on Computational Fluid Dynamics in the Oil & Gas, Metallurgical and Process Industries. – SINTEF Academic Press, 2017.