Investigation of discretization uncertainty in Monte Carlo neutron transport simulations of the Molten Salt Fast Reactor (MSFR)
DOI:
https://doi.org/10.15392/2319-0612.2023.1317Keywords:
MSFR, Neutronics, Monte Carlo, Extended GCI, Mesh based simulationAbstract
In the present work, an assessment of the Neutronic Benchmark of the Molten Salt Fast Reactor (MSFR) was performed using mesh based Monte Carlo Neutron Transport (MCNT) calculations with numerical uncertainty quantification due to discretization in neutronic parameters. Calculations with Constructive Solid Geometry (CSG) models where made as a baseline for the developed mesh based models. The numerical uncertainty given by the mesh utilization is evaluated using an extended version of the Grid Convergence Index (GCI). The fuel salt reprocessing is evaluated regarding a constant reprocessing rate. The fuel salt inventory variation with time for the developed models (CSG and meshed) is presented. The differences caused by the discretization procedure are noticeable, which shows that mesh based MCNT require careful mesh sensitivity evaluation and further validation.
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NEGRI, O. Modelling of neutronics and thermal-hydraulics of molten salt reactors. Ph.D. thesis, University of Manchester, 2020.
MERLE-LUCOTTE, E; HEUER, D; ALLIBERT, M; BROVCHENKO, M; CAPELLAN, N; GHETTA, V. Launching the thorium fuel cycle with the molten salt fast reactor. Proceedings of ICAPP, v. 2, 2011.
BROVCHENKO, M; KLOOSTERMAN, J.-L; LUZZI, L; MERLE, E; HEUER, D; LAUREAU, A; FEYNBERG, Olga; IGNATIEV, V; AUFIERO, M; CAMMI, A; et al. Neutronic benchmark of the molten salt fast reactor in the frame of the EVOL and MARS collaborative projects. EPJ Nuclear Sci. Technol., v. 5, p. 2, 2011.
VIEIRA, T; BARROS, G; CAMPOLINA, D; VASCONCELOS, V; DOS SANTOS, A. Study of a fine-mesh 1: 1 Computational Fluid Dynamics--Monte Carlo neutron transport coupling method with discretization uncertainty estimation. Annals of Nuclear Energy, v. 148, p. 107718, 2020.
LEPPÄNEN, J; VALTAVIRTA, V; VIITANEN, T; AUFIERO, M. Unstructured mesh based multi-physics interface for CFD code coupling in the Serpent 2 Monte Carlo code, Technical Report, 2015.
LEPPÄNEN, J; AUFIERO, M. Development of an unstructured mesh based geometry model in the Serpent 2 Monte Carlo Code, Technical Report, 2015.
HANSEN, G; OWEN, S. Mesh generation technology for nuclear reactor simulation; barriers and opportunities, Nuclear Engineering and Design, v. 238, p. 2590-2605, 2008.
DAWN, W; PALMTAG, S. Solving the Neutron Transport Equation for Microreactor Modeling Using Unstructured Meshes and Exascale Computing Architectures, Available at SSRN 4045437, 2022.
KNUPP, P; Remarks on Mesh Quality., Technical Report, Sandia National Lab.(SNL-NM), Albuquerque, NM (United States), 2007.
LAUREAU, A; HEUER, D; MERLE-LUCOTTE, E; RUBIOLO, P; ALLIBERT, M; AUFIERO, M. Transient coupled calculations of the Molten Salt Fast Reactor using the transient fission matrix approach, Nuclear Engineering and Design, v. 316, p. 112-124, 2017.
KELLEY, C; BERNHOLC, J; BRIGGS, E; HAMILTON, S; L, L; YANG, C. Mesh independence of the generalized Davidson algorithm, Journal of Computational Physics, v. 409, p. 109322, 2020.
JARETEG, K; VINAI, P; DEMAZIÈRE, C. Fine-mesh deterministic modeling of PWR fuel assemblies: Proof-of-principle of coupled neutronic/thermal--hydraulic calculations, Annals of Nuclear Energy, v. 68, p. 247-256, 2014.
GASTON, D; FORGET, B; SMITH, K; HARBOUR, L; RIDLEY, G; GIUDICELLI, G. Method of Characteristics for 3D, Full-Core Neutron Transport on Unstructured Mesh, Nuclear Technology, v. 207, p. 931-953, 2021.
LEPPÄNEN, J; PUSA, M; VIITANEN, T; VALTAVIRTA, V; KALTIAISENAHO, T. The Serpent Monte Carlo code: Status, development and applications in 2013, Annals of Nuclear Energy, v. 82, p. 142-150, 2015. Joint International Conference on Supercomputing in Nuclear Applications and Monte Carlo 2013, SNA + MC 2013. Pluri- and Trans-disciplinarity, Towards New Modeling and Numerical Simulation Paradigms.
GEUZAINE, C; REMACLE, J.-F. Gmsh: A 3-D finite element mesh generator with built-in pre-and post-processing facilities, International journal for numerical methods in engineering, v. 79, p. 1309-1331, 2009.
CELIK, I; GHIA, U; ROACHE, P; FREITAS, C; COLOMAN, H; RAAD, P. Procedure of Estimation and Reporting of Uncertainty Due to Discretization in CFD Applications, J. Fluids Eng., v. 130, p. 078001, 2008.
ROACHE, P. Fundamentals of Verification and Validation, Hermosa Publishers, 2009.
REZENDE, H; SANTOS, A; NAVARRO, M; JORDÃO, E. Verification and Validation of a thermal stratification experiment CFD simulation, Nuclear Engineering and Design, v. 248, p. 72-81, 2012.
DOS SANTOS, A. Investigação Numérica e Experimental do Escoamento de água em feixe de varetas representativo de elementos combustíveis nucleares de reatores do tipo PWR, Ph.D. thesis, Universidade Federal de Minas Gerais, Departamento de Engenharia Mecânica, 2012.
MATOZINHOS, C; DOS SANTOS, A. Two-phase CFD simulation of research reactor siphon breakers: A verification, validation and applicability study, Nuclear Engineering and Design, v. 326, p. 7-16, 2018.
DOLIGEZ, X; HEUER, D; MERLE-LUCOTTE, E; ALLIBERT, M; GHETTA, V. Coupled study of the Molten Salt Fast Reactor core physics and its associated reprocessing unit, Annals of Nuclear Energy, v. 64, p. 430-440, 2014.
WHEELER, A; CHVÁLA, O. Molten Salt Reactor Sourdough Refueling and Waste Management Strategy, Journal of Nuclear Engineering, v. 2, p. 471-483, 2021.
ASHRAF, O; TIKHOMIROV, G. Preliminary study on the online reprocessing and refueling scheme for SD-TMS reactor, Journal of Physics: Conference Series, v. 1439, p. 012005, IOP Publishing, 2020.
ASHRAF, O; RYKHLEVSKII, A; TIKHOMIROV, G; HUFF, K. Whole core analysis of the single-fluid double-zone thorium molten salt reactor (SD-TMSR), Annals of Nuclear Energy, v. 137, p. 107115, 2020.
RYKHLEVSKII, A; BAE, J; HUFF, K. Modeling and simulation of online reprocessing in the thorium-fueled molten salt breeder reactor, Annals of Nuclear Energy, v. 128, p. 366-379, 2019.
FIORINA, C; AUFIERO, M; CAMMI, A; FRANCHESCHINI, F; KREPEL, J; LUZZI, L; MIKITYUK, K; RICOTTI, M. Investigation of the MSFR core physics and fuel cycle characteristics, Progress in Nuclear Energy, v. 68, p. 153-168, 2013.
CHADWICK, M; HERMAN, M; OBLOZINSK, P; DUNN, M; DANON, Y; KAHLER, A; SMITH, D; PRITYCHENKO, B; ARBANAS, G; ARCILLA, R; et al. ENDF/B-VII. 1 nuclear data for science and technology: cross sections, covariances, fission product yields and decay data, Nuclear data sheets, v. 112, p. 2887-2996, 2011.
AUFIERO, M; BROVCHENKO, M; CAMMI, A; CLIFFORD, I; GEOFFROY, O; HEUER, D; LAUREAU, A; LOSA, M; LUZZI, L; MERLE-LUCOTTE, E; et al. Calculating the effective delayed neutron fraction in the molten salt fast reactor: analytical, deterministic and Monte Carlo approaches, Annals of Nuclear Energy, v. 65, p. 78-90, 2014.
AUFIERO, M. Development of advanced simulation tools for circulating fuel nuclear reactors, Ph.D. thesis, Politecnico di Milano, 2014.
AUFIERO, M; CAMMI, A; FIORINA, C; LEPPÄNEN, J; LUZZI, L; RICOTTI, M. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor, Journal of Nuclear Materials, v. 441, p. 473-486, 2013.
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