Kazakhstan's National Nuclear Center Introduces Computer Modeling Methods for Reactor Safety

The National Nuclear Center of the Republic of Kazakhstan (NNC RK) recently presented its modern computer modeling methods used in the safety justification of reactor facilities. Arthur Suraev, a leading researcher at the laboratory of the branch Institute of Atomic Energy, stated that high-precision computational models have become an important tool for assessing reactor normal operation, design-basis accidents, and beyond-design-basis accident conditions.

Suraev noted that computer modeling enables preliminary evaluation of neutron physics, thermal-hydraulic, and structural response processes inside a reactor before conducting complex and expensive experiments. By analyzing interrelated physical processes, researchers can identify potential changes during the design and research stages and optimize safety-related technical solutions accordingly.

Currently, NNC RK uses licensed software such as MCNP6.3 and ANSYS Academic Research for computational support of reactor experiments. MCNP6.3, based on the Monte Carlo method and continuous-energy nuclear data libraries, simulates the transport of neutrons and gamma quanta in complex three-dimensional geometries and is used to calculate the neutron multiplication factor, spatial distribution of neutron flux, radiation heating, and local energy release in nuclear fuel.

ANSYS is primarily used for simulating macroscopic physical processes, including coolant flow calculations, single-phase and multiphase flow analysis, temperature distribution determination, and assessment of thermal stress and mechanical strength of fuel elements and fuel assemblies under different operating modes. The software is capable of reflecting the behavior of reactor structures under thermal and mechanical loads.

The coupled use of the two software packages improves computational accuracy. Typically, MCNP first calculates the energy release distribution in the core, and the data is then transferred to ANSYS to compute the temperature field; subsequently, the updated temperature information is fed back to MCNP for subsequent calculations that account for temperature feedback effects. Through this data exchange, researchers can build high-precision digital twin models of the core and more reliably predict nuclear fuel behavior under normal operation and accident conditions.

Suraev also emphasized that computational methods cannot be separated from experimental validation. NNC RK uses the IVG.1M and IGR research reactors, as well as specialized out-of-pile experimental facilities, to validate computational models. Experimental results are used to refine mathematical models, thereby improving the accuracy and credibility of computational results.

Validated computational methods can be applied to analyze the performance of fuel elements and fuel assemblies in power reactors such as VVER and PWR, including assessing the limits of maintaining fuel rod cladding integrity under accident conditions, simulating fuel degradation and melting processes, studying interactions between fuel and structural materials, and predicting the performance of next-generation accident-tolerant fuels.

NNC RK believes that the combination of modern computer modeling and experimental platforms can form a scientific and technical system for the safety justification of reactor facilities, helping to improve computational reliability, reduce operational risks, and provide support for the development and improvement of reactor technologies.