Liquid metal thermal hydraulics R&D at European scale: achievements and prospects

Abstract

A significant role for a future nuclear carbon-free energy production is attributed to fast reactors, mostly employing a liquid metal as a coolant. This paper summarizes the efforts that have been undertaken in collaborative projects sponsored by the European Commission in the past 20 years in the fields of liquid-metal heat transfer modeling, fuel assembly and core thermal hydraulics, pool and system thermal hydraulics, and establishment of best practice guidelines and verification, validation, and uncertainty quantification (UQ). The achievements in these fields will be presented along with the prospects on topics which will be studied collaboratively in Europe in the years to come. These prospects include further development of heat transfer models for applied computational fluid dynamics (CFD), further analysis of the consequences of fuel assembly blockages on coolant flow and temperature, analysis of the thermal hydraulic effects in deformed fuel assemblies, extended validation of three-dimensional pool thermal hydraulic CFD models, and further development and validation of multi-scale system thermal hydraulic methods.

Graphical abstract

Keywords

liquid metal / thermal hydraulics / Europe

References

[1] BP. Energy Outlook 2019 edition. 2019

[2] Capros P, De Vita A, Tasios N, EU reference scenario 2016—energy, transport and GHG emissions—trends to 2050. Luxembourg: Publications Office of the European Union, 2016

[3] Quental N, Buttle D, Abrar S, Strategic energy technology (SET) plan. Luxembourg: Publications Office of the European Union, 2017

[4] International Energy Agency. World energy outlook 2018. 2018

[5] International Atomic Energy Agency (IAEA). Energy, Electricity and Nuclear Power Estimates for the Period up to 2050. 2010 edition. Vienna: International Atomic Energy Agency, 2010

[6] Lassiter J B. The Future of Nuclear Energy in a Carbon-Constrained World: An Interdisciplinary MIT Study. Cambridge: MIT Energy Initiative, 2018

[7] OECD/NEA. The costs of decarbonisation: system costs with high shares of nuclear and renewables. NEA No. 7299, 2019

[8] Edenhofer O, Pichs-Madruga R, Sokona Y. Climate Change 2014: Mitigation of Climate Change. Working Group III Contribution to the Fifth Assessment Report of the Intergovernmental Panel on Climate Change. New York: Cambridge University Press, 2014

[9] Nuclear Energy Agency, Organisation for Economic Co-operation and Development. Uranium 2018: resources, production and demand. NEA No. 7413, 2018

[10] World Nuclear Association. Information library. 2020–09–16

[11] Roelofs F. Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors. Duxford, UK: Woodhead Publishing, Elsevier, 2018

[12] IAEA. Catalogue of facilities in support of LMFNS. 2020–09–16

[13] Pfrang W, Struwe D. Assessment of correlations for heat transfer to the coolant for heavy liquid metal cooled core designs. Forschungzentrum Karlsruhe Report, FZKA 7352, 2007

[14] Mikityuk K. Heat transfer to liquid metal: review of data and correlations for tube bundles. Nuclear Engineering and Design, 2009, 239(4): 680–687

[15] Jaeger W. Heat transfer to liquid metals with empirical models for turbulent forced convection in various geometries. Nuclear Engineering and Design, 2017, 319: 12–27

[16] OECD/NEA. Handbook on Lead-Bismuth Eutectic Alloy and Lead Properties, Materials Compatibility, Thermal-hydraulics and Technologies. Paris: OECD Publishing, 2015

[17] Passerini S, Gerardi C, Grandy C, IAEA NAPRO coordinated research project: physical properties of sodium—overview of the reference database and preliminary analysis results. In: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (2017), Yekaterinburg, Russia, 2017

[18] Shams A, Roelofs F, Tiselj I, A collaborative effort towards the accurate prediction of turbulent flow and heat transfer in low-Prandtl number fluids. Nuclear Engineering and Design, 2020, 366: 110750

[19] Roelofs F. Liquid metal thermal hydraulics: state-of-the-art and future perspectives. Nuclear Engineering and Design, 2020, 362: 110590

[20] Shams A, De Santis A, Roelofs F. An overview of the AHFM-NRG formulations for the accurate prediction of turbulent flow and heat transfer in low-Prandtl number flows. Nuclear Engineering and Design, 2019, 355: 110342

[21] Roelofs F, Dovizio D, Uitslag-Doolaard H, Core thermal hydraulic CFD support for liquid metal reactors. Nuclear Engineering and Design, 2019, 355: 110322

[22] Roelofs F, Uitslag-Doolaard H, Mikuz B, CFD and experiments for wire-wrapped fuel assemblies. In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH–18), Portland, USA, 2019, 5716–5729

[23] Kennedy G, Van Tichelen K, Pacio J, Thermal-hydraulic experimental testing of the MYRRHA wire-wrapped fuel assembly. Nuclear Technology, 2020, 206(2): 179–190

[24] Pacio J, Daubner M, Wetzel T, Inter-wrapper flow: LBE experiments and simulations. In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH –18), Portland, USA, 2019, 796–809

[25] Roelofs F, Gopala V, Jayaraju S, Review of fuel assembly and pool thermal hydraulics for fast reactors. Nuclear Engineering and Design, 2013, 265: 1205–1222

[26] Grishchenko D, Jeltsov M, Kööp K, The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes. Nuclear Engineering and Design, 2015, 290: 144–153

[27] Jeltsov M, Grishchenko D, Kudinov P. Validation of Star-CCM+ for liquid metal thermal-hydraulics using TALL-3D experiment. Nuclear Engineering and Design, 2019, 341: 306–325

[28] Martelli D, Tarantino M, Forgione N. CIRCE-ICE PLOHS experimental campaign. Nuclear Engineering and Design, 2019, 355: 110307

[29] Zwijsen K, Dovizio D, Moreau V, CFD modelling of the CIRCE facility. Nuclear Engineering and Design, 2019, 353: 110277

[30] Van Tichelen K, Mirelli F, Greco M, E-SCAPE: a scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations. Nuclear Engineering and Design, 2015, 290: 65–77

[31] Visser D, Keijers S, Lopes S, CFD analyses of the European scaled pool experiment E-SCAPE. Nuclear Engineering and Design, 2020, 358: 110436

[32] Lopes S, Koloszar L, Planquart P, Hunting for the correct pressure drop in a scaled reactor pool: effect of geometry, mesh resolution, turbulence model and mass flow. In: International Topical Meeting on Advances in Thermal Hydraulics (ATH’2020), Paris, France, 2020

[33] Frignani M, Alemberti A, Tarantino M, ALFRED staged approach. In: International Congress on Advances in Nuclear Power Plants (ICAPP 2019), Juan-les-pins, France, 2019

[34] SCK CEN. MYRRHA: multi-purpose hybrid research reactor for high-tech applications: a research infrastructure for a new era. 2020–09–16

[35] Moreau V, Profir M, Alemberti A, Pool CFD modelling: lessons from the SESAME project. Nuclear Engineering and Design, 2019, 355: 110343

[36] Moreau V, Profir M, Keijers S, An improved CFD model for a MYRRHA based primary coolant loop. Nuclear Engineering and Design, 2019, 353: 110221

[37] Koloszar L, Planquart P, Van Tichelen K, Numerical simulation of loss-of-flow transient in the MYRRHA reactor. Nuclear Engineering and Design, 2020, 363: 110675

[38] Bestion D. System thermalhydraulics for design basis accident analysis and simulation: status of tools and methods and direction for future R&D. Nuclear Engineering and Design, 2017, 312: 12–29

[39] Uitslag-Doolaard H, Alcaro F, Roelofs F, Multiscale modelling of the PHENIX dissymmetric test benchmark. Nuclear Engineering and Design, 2020, 356: 110375

[40] IAEA. Benchmark analysis of FFTF loss of flow without scram test. 2020-9–16

[41] Li S, Gerschenfeld A, Sageaux T. Onset of natural convection in a sodium-cooled fast reactor during a station black-out: blind benchmark of safety assessment using multi-scale coupled thermal- hydraulics codes. In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH–18), Portland, USA, 2019, 4418–4429

[42] Toti A, Vierendeels J, Belloni F. Improved numerical algorithm and experimental validation of a system thermal-hydraulic/CFD coupling method for multi-scale transient simulations of pool-type reactors. Annals of Nuclear Energy, 2017, 103: 36–48

[43] Gerschenfeld A. Multiscale and multiphysics simulation of sodium fast reactors: from model development to safety demonstration. In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH–18), Portland, USA, 2019, 4164–4177

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