Nuclear Thermal-hydraulic Research Group
Research keywords:
· CFD
· thermal-hydraulics
· nuclear safety analysis
· nuclear
· APROS
· system code
Horizon Europe keywords:
· Clean, sustainable, secure and competitive energy supply
Introduction of the Research Group
The Nuclear Thermo-hydraulic Research Group has decades of extensive experience in studying thermohydraulic phenomena in various parts of the nuclear power plants currently in operation or under construction, as well as in the ITER and DEMO fusion facilities. In our research, we examine in detail the coolant mixing and temperature distribution inside the reactor pressure vessel and fuel assemblies with the ANSYS CFX code; the applicability of supercritical pressure water to nuclear power plants; cooling of the cluster of spent fuel storage tubes with a natural airflow; the use of helium as a coolant in various fusion equipment. The CFD simulations are mainly performed for the VVER-440 and VVER-1200 type reactors, but we also deal with the investigation of 4th generation reactors (SCWR, molten-salt, gas-cooled). The research group uses the measurement results of the PIV laboratory located at BME NTI to validate the CFD models. Thermo-hydraulic analysis of single- and two-phase processes in the complex systems of nuclear power plants is performed using different system codes (APROS, TRACE and RELAP5). BSc, MSc and PhD students are also involved in solving our tasks.
Watch our 3-minute introductory video:
Achievements
- Significant reduction of helium pressure drop inside the HCLL DEMO helium manifold system by redesigning its inner structure.
- Validation of the CFD model of the VVER-440/213 reactor pressure vessel with different measurement results
- Validation of the CFD model of VVER-440/213 type fuel assemblies.
- Development of a validated CFD model of the full cross-section of the ALLEGRO gas-cooled fast reactor ceramic assembly. The model describes all the structural elements of the active part of the assembly, their thermal conductivity and the heat transfer in the coolant and the gas gap.
- Development of a validated CFD model for the study of mixing vanes in the ceramic assembly of the ALLEGRO gas-cooled fast reactor. Development of a test methodology to investigate the effectiveness of different mixing vane types, considering heat transfer and mixing coefficients as well as the pressure drop generated by the vanes.
- Acquisition of competence for thermohydraulic investigation of supercritical pressure water.
- Active participation in various international projects (ECC-SMART, SafeG, etc.)
Publications
T. Varju, Á. Aranyosy, R. Orosz, V. Holl, T. Hajas, A. Asódi; Analysis of the IAEA SPE-4 small-break LOCA experiment with RELAP5, TRACE and APROS system codes; Nuclear Engineering and Design; 2021
GI Orosz, S. Tóth, A. Asódi; Detailed thermal modeling of the ALLEGRO ceramic assembly, Nuclear Engineering and Design; 376; 2021
G. Zsíros, S. Tóth, A. Asódi; Analysis of coolant flow in central tube of VVER-440 fuel assemblies; Core technology; 2012
A. Kiss; T. Cutter; A. Asódi, Numerical analysis on inlet and outlet sections of a test fuel assembly for a Supercritical Water Reactor, Nuclear Engineering and Design; 295; 2015
S. Tóth, A. Asódi; Determination of mixing factors for VVER-440 fuel assembly head
Nuclear Engineering and Design; 264; 2013
Journals
Nuclear Engineering and Design
Annals of Nuclear Energy
Kerntechnik
Journal of Nuclear Engineering and Radiation Science
Infrastructure
PIV and TRATEL labs
CFD lab
computer cluster
CFX, APROS and NRC license
Projects
EUROfusion Horizon 2020 and Europe, Thermo-hydraulic analysis of different components of EU HCLL and HCPB DEMO concepts, 2014-2027, EUROfusion
Development of a Paks NPP specific VVER-440/213 primary and secondary APROS6 model and thermal-hydraulic analysis of Paks NPP in shutdown states using APROS code; 2017 – 2022; Paks NPP
Joint European Canadian Chinese development of Small Modular Reactor Technology, 2020-2024, European Commission (RTD/D/04)
Safety of GFR through innovative materials, technologies and processes; 2020-2024, European Commission
SafeG (NUMBER 945041), 2020 - 2024, European Atomic Energy Community
Industry relations
Paks NPP.
Paks II. NPP.
Hungarian Atomic Energy Authority
International Atomic Energy Agency
SOM System Kft.
Conferences
Apros User Group Seminar; Helsinki/Espoo (Finland); Tamás Varjú (2022, 2019, 2017); Bendegúz Kopp (2022, 2019); participant/presenter
29th Symposium of AER on VVER Reactor Physics and Reactor Safety; Energoland, Mochovce NPP, Slovakia; Gergely Imre Orosz, Sándor Tóth, Attila Asódi, 2019; participant/presenter
10th International Symposium on Supercritical Water-Cooled Reactors; Prague, Czech Republic; 2021; Organizer