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Modernization of the SC-INT subchannel thermalhydraulic code
(НИЯУ МИФИ, 2026) Vertikov, E. A.; Zaporzhin, K. V.; Oleksyuk, D. A.; Khamaza, V. A.; Khudykin, A. M.; Glazov, M. A.; Morozkin, O. N.
The article presents the main results of the work cycle on modernizing the source code of the SC-INT computer program designed for subchannel thermal-hydraulic calculations of the water-cooled nuclear reactors cores. The mathematical description of the program is briefly provided, including the method of allocation of control volumes in space, the discrete analog of the basic conservation laws forming the system of nonlinear equations, as well as the method of its solution. The path passed on the internal modernization of the program is described in detail: ejection of outdated Fortran programming language constructions, transition to structure-oriented approach of writing source code, development of modular architecture, as well as implementation of the alternative numerical algorithm for solving the main system of nonlinear equations using the PETSc library. As an example of the SC-INT program capabilities, which appeared after the above described modernizations, the results of thermal-hydraulic calculation in fine-mesh subchannel approximation of a full-scale VVER-1000 reactor core are presented. The core under consideration is assembled from fuel assemblies of different designs: with and without installed «Vikhr» and «Progonka» type intensifier grids. It is demonstrated that the residuals on the main coolant parameters achieved in the simulation of the full-scale core match in order with the corresponding values characteristic for calculations of small-scale experimental fuel assembly models. Thermal-hydraulic calculations of full-scale cores in the subchannel approximation opens the possibility for development of coupled program complexes designed for improved estimation of the parameters of multiphysics processes in the cores of water-cooled nuclear reactors.
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Pretreatment of radioactive concentrate from leadbismuth cooled fast reactors: flocculant screening and process optimization
(НИЯУ МИФИ, 2026) Peng, S.; Wang, D.; Li, S.; Li, J.; Xie, H.
The processing of radioactive concentrates generated during the operation of lead bismuth fast reactors is extremely difficult, with complex components including heavy metals such as Pb and Bi, radioactive nuclides such as Cs, Sr, and Co, alpha emitters such as 210-Po, as well as organic matter and suspended particles. This study focuses on the pretreatment challenges of this radioactive liquid waste, systematically screening 13 types of mature commercial flocculants and evaluating their efficiency in the flocculative removal of key heavy metals and simulated radionuclides from the concentrate. The experimental results showed that the flocculant #9 had a removal efficiency of over 99% for Pb, Bi, and Co within 20 minutes, and a removal efficiency of 96% for the 210-Po analog element Te. The effluent pH was close to neutral, which was superior to other reagents. Through optimization via single-factor experiments and Response Surface Methodology, the optimal process parameters for flocculant #9 were determined as follows: pH 7.8, dosage of 1.17 g/L, and flocculation time of 15.9 minutes. This study provides an efficient and low-cost pretreatment process for radioactive concentrate from lead-bismuth reactors, which can significantly reduce the subsequent deep purification load of 21°Po and offers technical support for the optimization of nuclear waste liquid treatment processes.
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Nuclear Energy and Technology (NUCET)
2026 -12 - 1
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Nuclear Energy and Technology (NUCET)
Nuclear Energy and Technology (NUCET) (2026 -12)
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Safety-oriented CFD analysis of the PeLUIt-40 pebblebed HTGR under partial flow and LOCA scenarios
(НИЯУ МИФИ, 2026) Afif, M. T.; Sulistyo, F. Y.; Veretennikov, D. G.
In this study, the thermal–hydraulic behavior of the Indonesian Micro Reactor PeLUIt-40, a 40 MWt modular pebble-bed high-temperature gas-cooled reactor (HTGR), was analyzed under reduced coolant flow and loss-of-coolant accident (LOCA) conditions. Three-dimensional computational fluid dynamics (CFD) simulations using a porous-media core model and volumetric heat sources from neutronic analysis were performed. Steady-state simulations were conducted for helium flow rates from 100% to 25% of nominal, and a transient LOCA was simulated by reducing flow to zero over ten seconds. The results show that maximum core temperatures increased nonlinearly with reduced flow, exceeding TRISO fuel limits at 25% flow, while outlet duct temperatures remained well homogenized. During the LOCA, passive buoyancy-driven circulation limited temperature rise, stabilizing around 1026 °C. These findings provide preliminary insight into PeLUIt-40 thermal hydraulics, highlighting the need for further validation and extended transient analysis to confirm safety margins.