Home Technology Coupled AC2-CFD simulations for a high-pressure core melt accident scenario
Article
Licensed
Unlicensed Requires Authentication

Coupled AC2-CFD simulations for a high-pressure core melt accident scenario

  • Joachim Herb EMAIL logo , Hristo V. Hristov and Thomas Steinrötter
Published/Copyright: February 23, 2024
Become an author with De Gruyter Brill

Abstract

Even though very unlikely to occur, severe accident scenarios in nuclear power plants have to be analyzed. During high-pressure core meltdown scenarios in a pressurized water reactor the primary circuit should fail first. Previous analyses found that a free convection flow within the vertical steam generator (SG) tubes with a simultaneous stratified gas counterflow in the hot legs could arise. This phenomenon leads to higher thermal loads on individual SG tubes which might then fail leading to a containment bypass and the release of radioactive material into the environment. Lumped parameter system codes used for safety analyses do not provide the models necessary to simulate phenomena like mixing in three-dimensional flows and could not consider local turbulence effects. Computational fluid dynamic (CFD) codes provide such capabilities but are much more computationally expensive. Coupling of a system code with a CFD code can therefore be used to simulate such phenomena. The advantages of both approaches can be maximized by splitting up the simulation domain between the codes, depending on the expected flow conditions. The system code AC2 coupled with the CFD code OpenFOAM was used to simulate part of the severe accident transient. Free convection in the hot leg and the U-tubes of the vertical SG was observed in case of high-pressure severe accident sequences. The thermal load of individual SG tubes has been estimated from the results. These loads can be used as inputs for structural-mechanical analyses to estimate which part of the primary circuit would fail first.


Corresponding author: Joachim Herb, Safety Research Division, Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Botzmannstraße 14, 85748 Garching bei München, Germany, E-mail:

  1. Research ethics: Not applicable.

  2. Author contributions: The authors have accepted responsibility for the entire content of this manuscript and approved its submission.

  3. Competing interests: The authors state no conflict of interest.

  4. Research funding: This work was funded by the German Federal Ministry of the Environment, Nature Conservation, Nuclear Safety and Consumer Protection within the project 4719R01376 “Further research on the HD core meltdown process by coupling a CFD loop model with ATHLET-CD”.

  5. Data availability: Not applicable.

References

Band, S., Bläsius, C., Scheuerer, M., and Steinrötter, T. (2017). Thermohydraulisches Verhalten und Komponentenverhalten eines DWR bei ausgewähltem Kernschmelzszenarium infolge Station Blackout (SBO). GRS-473. Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Available at: https://www.grs.de/de/aktuelles/publikationen/grs-473-thermohydraulisches-verhalten-und-komponentenverhalten-eines-dwr (Accessed 08 January 2024).Search in Google Scholar

Boyd, C.F. and Armstrong, K.W. (2010). Computational fluid dynamics analysis of natural circulation flows in a pressurized-water reactor loop under severe accident conditions. NUREG-1922, Office of Nuclear Regulatory Research, Available at: https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1922/index.html (Accessed 08 January 2024).Search in Google Scholar

Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit (2022). Änderung der Bekanntmachung zu den „Sicherheitsanforderungen an Kernkraftwerke“ vom 25, Februar 2022. BAnz AT 15.03.2022 B3, Available at: https://www.base.bund.de/SharedDocs/Downloads/BASE/DE/rsh/3-bmub/3_0_1.pdf?__blob=publicationFile&v=%201 (Accessed 08 January 2024).Search in Google Scholar

CFD Direct Ltd. (2020). OpenFOAM user guide, version 8, Available at: https://doc.cfd.direct/openfoam/user-guide-v8/index (Accessed 08 January 2024).Search in Google Scholar

Dittus, F.W. and Boelter, L.M.K. (1930). Heat transfer in automobile radiators of the tubular type. Univ. Calif. Publ. Eng. 2: 443–461.Search in Google Scholar

Fiorina, C., Clifford, I., Kelm, S., and Lorenzi, S. (2021a). On the development of multi-physics tools for nuclear reactor analysis based on OpenFOAM®: state of the art, lessons learned and perspectives. Nucl. Eng. Des. 111604, https://doi.org/10.1016/j.nucengdes.2021.111604.Search in Google Scholar

Fiorina, C., Shriwise, P., Dufresne, A., Ragusa, J., Ivanov, K., Valentine, T., Lindley, B., Kelm, S., Shwageraus, E., Monti, S., et al.. (2021b). An initiative for the development and application of open-source multi-physics simulation in support of R&D and E&T in nuclear science and technology. EPJ Web Conf. 247, 2040, https://doi.org/10.1051/epjconf/202124702040.Search in Google Scholar

Fletcher, C.D., Beaton, R.M., Palazov, V.V., Caraher, D.L., and Shumway, R.W. (2010). SCDAP/RELAP5 thermal-hydraulic evaluations of the potential for containment bypass during extended station blackout severe accident sequences in a Westinghouse four-loop PWR. NUREG/CR-6995. Office of Nuclear Regulatory Research, Available at: https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6995/index.html (Accessed 08 January 2024).Search in Google Scholar

Greenshields, C.J. and Weller, H.G. (2022). Notes on computational fluid dynamics. General principles. CFD Direct Limited, Reading, Available at: https://doc.cfd.direct/notes/cfd-general-principles/ (Accessed 08 January 2024).Search in Google Scholar

Herb, J. (2014). Coupling of OpenFOAM with thermo-hydraulic simulation code ATHLET. 9th OpenFOAM® Workshop, Zagreb.Search in Google Scholar

Herb, J. (2019). Coupled OpenFOAM-ATHLET simulations of the primary circuit of a liquid sodium cooled reactor. 14th OpenFOAM Workshop, Duisburg, Available at: https://openfoam-extend.sourceforge.net/OpenFOAM_Workshops/OFW14_2019_Duisburg/www_dot_conftool_dot_com/ofw14/index.php/Herb-Coupled_OpenFOAM-ATHLET_simulations_of_the_primary_circuit_of_a_liquid_sodium_cooled_reactor-115.pdf@page=downloadPaper&filename=Herb-Coupled_OpenFOAM-ATHLET_simulations_of_the_primary_circuit_of_a_ (Accessed 08 January 2024).Search in Google Scholar

Herb, J. (2021). Coupled ATHLET-OpenFOAM simulation of a station blackout for the ESFR-SMART reactor concept. German CFD Network of Competence, Online.Search in Google Scholar

Herb, J., Hristov, H., and Steinrötter, T. (2023). Weiterführende Forschungsarbeiten zum HD-Kernschmelzablauf in einem DWR durch Kopplung eines CFD-Schleifenmodells mit ATHLET-CD. GRS-678. Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Available at: https://www.grs.de/de/aktuelles/publikationen/grs-678.Search in Google Scholar

Hicken, E.F. and Scholl, K.H. (1985). Blowdown tests to verify calculation methods for loss-of-coolant accidents. Nuclear Engineering and Design 88: 277–286, https://doi.org/10.1016/0029-5493(85)90163-3.Search in Google Scholar

Hristov, H. and Herb, J. (2019). On the development of coupled code ATHLET-OpenFOAM solvers for safety related analyses in nuclear facilities. 14th OpenFOAM Workshop, Duisburg, Available at: https://openfoam-extend.sourceforge.net/OpenFOAM_Workshops/OFW14_2019_Duisburg/www_dot_conftool_dot_com/ofw14/index.php/Hristov-On_the_development_of_coupled_code_ATHLET-OpenFOAM_solvers-114.pdf@page=downloadPaper&filename=Hristov-On_the_development_of_coupled_code_ATHLET-OpenFOAM_solvers-114.pdf&form_id=114&form_version= (Accessed 08 January 2024).Search in Google Scholar

Iglesias Moreno, A. (2017). Implementation of an advanced numerical method for the optimization of the ATHLET-OpenFOAM coupling capabilities, Master thesis. Technische Universität München (TUM), Garching bei München.Search in Google Scholar

Issa, R. (1986). Solution of the implicitly discretised fluid flow equations by operator-splitting. J. Comp. Phys. 62: 40–65, https://doi.org/10.1016/0021-9991(86)90099-9.Search in Google Scholar

Kelm, S., Kampili, M., Liu, X., George, A., Schumacher, D., Druska, C., Struth, S., Kuhr, A., Ramacher, L., Allelein, H.-J., et al.. (2021). The tailored CFD package ‘containmentFOAM’ for analysis of containment atmosphere mixing, H2/CO mitigation and aerosol transport. Fluids 6: 100, https://doi.org/10.3390/fluids6030100.Search in Google Scholar

Lehnigk, R., Bruschewski, M., Huste, T., Lucas, D., Rehm, M., and Schlegel, F. (2023). Sustainable development of simulation setups and addons for OpenFOAM for nuclear reactor safety research. Kerntechnik 88: 131–140, https://doi.org/10.1515/kern-2022-0107.Search in Google Scholar

Lemmon, E.W., Bell, I.H., Huber, M.L., and McLinden, M.O. (2021). Thermophysical Properties of fluid systems. NIST Chemistry WebBook, NIST standard reference database 69. NIST Chemistry Webbook, SRD69. National Institute of Standards and Technology, Available at: https://webbook.nist.gov/chemistry/fluid/ (Accessed 08 January 2024).Search in Google Scholar

Menter, F.R. and Esch, T. (2001) Elements of industrial heat transfer prediction. 16th Brazilian Congress of mechanical engineering, Uberlandia.Search in Google Scholar

OECD Nuclear Energy Agency (NEA) (2015). Best practice guidelines for the use of CFD in nuclear reactor safety application – revision. NEA/CSNI/R(2014)11, Available at: https://www.oecd-nea.org/jcms/pl_19548/best-practice-guidelines-for-the-use-of-cfd-in-nuclear-reactor-safety-applications-revision (Accessed 08 January 2024).Search in Google Scholar

Papukchiev, A., Lerchl, G., Waata, C., and Frank, T. (2009) Extension of the simulation capabilities of the 1D system code ATHLET by coupling with the 3D software package ANSYS CFX. NURETH-13: Proceedings of the 13th international topical meeting on nuclear reactor thermal hydraulics.Search in Google Scholar

Patankar, S.V. and Spalding, D.B. (1972). A calculation procedure for heat, mass and momentum transfer in three-dimensional parabolic flows. Int. J. Heat Mass Transfer 15: 1787–1806, https://doi.org/10.1016/0017-9310(72)90054-3.Search in Google Scholar

Sancaktar, S., Salay, M., Iyengar, R., Azarm, A., and Majumdar, S. (2018). Consequential SGTR analysis for Westinghouse and Combustion engineering plants with thermally treated alloy 600 and 690 steam generator tubes. NUREG-2195. Office of Nuclear Regulatory Research, Available at: https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr2195/index.html (Accessed 08 January 2024).Search in Google Scholar

STF Solutions (2019). Solving for your own Sutherland Coefficients using Python, Available at: https://curiosityfluids.com/2019/04/24/solving-for-your-own-sutherland-coefficients-using-python/ (Accessed 31 Jul 2023).Search in Google Scholar

Toti, A., Vierendeels, J., and Belloni, F. (2017). Improved numerical algorithm and experimental validation of a system thermal-hydraulic/CFD coupling method for multi-scale transient simulations of pool-type reactors. Ann. Nucl. Energy 103: 36–48, https://doi.org/10.1016/j.anucene.2017.01.002.Search in Google Scholar

Wielenberg, A., Lovasz, L., Pandazis, P., Papukchiev, A., Tiborcz, L., Schöffel, P.J., Spengler, C., Sonnenkalb, M., and Schaffrath, A. (2019). Recent improvements in the system code package AC2 2019 for the safety analysis of nuclear reactors. Nucl. Eng. Des. 354: 110211, https://doi.org/10.1016/j.nucengdes.2019.110211.Search in Google Scholar

Received: 2023-10-06
Accepted: 2024-01-22
Published Online: 2024-02-23
Published in Print: 2024-04-25

© 2024 Walter de Gruyter GmbH, Berlin/Boston

Downloaded on 11.12.2025 from https://www.degruyterbrill.com/document/doi/10.1515/kern-2023-0106/pdf
Scroll to top button