Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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STRAUM-MATXST: A code system for multi-group neutron-gamma coupled transport calculation with unstructured tetrahedral meshes

  • MyeongHyeon Woo (Department of Nuclear Engineering, Hanyang University) ;
  • Ser Gi Hong (Department of Nuclear Engineering, Hanyang University)
  • Received : 2022.01.21
  • Accepted : 2022.07.03
  • Published : 2022.11.25
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Abstract

In this paper, a new multi-group neutron-gamma transport calculation code system STRAUM-MATXST for complicated geometrical problems is introduced and its development status including numerical tests is presented. In this code system, the MATXST (MATXS-based Cross Section Processor for SN Transport) code generates multi-group neutron and gamma cross sections by processing MATXS format libraries generated using NJOY and the STRAUM (SN Transport for Radiation Analysis with Unstructured Meshes) code performs multi-group neutron-gamma coupled transport calculation using tetrahedral meshes. In particular, this work presents the recent implementation and its test results of the Krylov subspace methods (i.e., Bi-CGSTAB and GMRES(m)) with preconditioners using DSA (Diffusion Synthetic Acceleration) and TSA (Transport Synthetic Acceleration). In addition, the Krylov subspace methods for accelerating the energy-group coupling iteration through thermal up-scatterings are implemented with new multi-group block DSA and TSA preconditioners in STRAUM.

Keywords

Acknowledgement

This work was supported by the NRF (National Research Foundation of Korea) through Project No. NRF-2019M2D2A1A02057890 and the Nuclear Safety Research Program through the Korea Foundation Of Nuclear Safety (KoFONS) using the financial resource granted by the Nuclear Safety and Security Commission (NSSC) of the Republic of Korea. (No. 2101075)

References

  1. T.A. Wareing, J.M. McGhee, J.E. Morel, ATTILA: a three-dimensional, unstructured tetrahedral mesh discrete ordinates transport code, Trans. Am. Nucl. Soc. 75 (1996). 
  2. R.L. Martz, The MCNP6 book on unstructured mesh geometry: user's guide, Los Alamos Natl. Lab. (2014) 1-77. LA-UR-11-05668. 
  3. T.D. Long, S.G. Hong, Implementation and Verification of adjoint neutron transport calculation in MUST code, in: Transactions of the Korea Nuclear Society Virtual Spring Meeting, Republic of Korea, 2020. July 9-10. 
  4. T. Wareing, J.M. McGhee, J. Morel, S. Pautz, Discontinuous finite element SN methods on three-dimensional unstructured grids, Nucl. Sci. Eng. 138 (2001) 256-268, https://doi.org/10.13182/NSE138-256. 
  5. S.G. Hong, Two subcell balance methods for solving the multigroup discrete ordinates transport equation with tetrahedral meshes, Nucl. Sci. Eng. 173 (2013), https://doi.org/10.13182/NSE11-38. 
  6. H. Muhammad, S.G. Hong, A three-dimensional Fourier analysis of fine mesh rebalance acceleration of linear discontinuous sub-cell balance method for diffusion equation on tetrahedral meshes, Ann. Nucl. Energy 133 (2019) 145-153, https://doi.org/10.1016/J.ANUCENE.2019.05.020. 
  7. Y. Azmy, E. Sartori, Nuclear Computational Science: a Century in Review, Springer, 2010. 
  8. R.N. Slaybaugh, Vermaak, T.M. Evans, G.G. Davidson, P.P.H. Wilson, Multigrid in energy preconditioner for Krylov solvers, J. Comput. Phys. 242 (2013) 405-419.  https://doi.org/10.1016/j.jcp.2013.02.012
  9. Christophe Geuzaine, Jean-Francois Remacle, Gmsh, Available at: http://gmsh.info/, 2020. 
  10. M.H. Woo, T.D. Long, S.G. Hong, Development of multi-group cross section processing program for MUST unstructured discrete ordinate transport code, in: Transactions of the Korea Nuclear Society Virtual Autumn Meeting, Republic of Korea, 2020. December 17-18. 
  11. T.W. Huang, D.L. Lin, C.X. Lin, Y. Lin, Taskflow: a lightweight parallel and heterogeneous task graph computing system, IEEE Trans. Parallel Distrib. Syst. 33 (2022), https://doi.org/10.1109/TPDS.2021.3104255. 
  12. Y. Saad, M.H. Schultz, GMRES: a generalized minimal residual algorithm for solving nonsymmetric linear systems, SIAM J. Sci. Stat. Comput. 7 (1986), https://doi.org/10.1137/0907058. 
  13. H.A. van der Vorst, Bi-CGSTAB: a fast and smoothly converging variant of Bi-CG for the solution of nonsymmetric linear systems, SIAM J. Sci. Stat. Comput. 13 (1992), https://doi.org/10.1137/0913035. 
  14. H. Muhammad, S.G. Hong, Diffusion synthetic acceleration with the fine mesh rebalance of the subcell balance method with tetrahedral meshes for SN transport calculations, Nucl. Eng. Technol. 52 (2020), https://doi.org/10.1016/j.net.2019.08.021. 
  15. G. Guennebaud, B. Jacob, Eigen v3, others, Available at: https://eigen.tuxfamily.org/index.php?title=Main_Page, 2010. 
  16. M.L. Adams, E.W. Larsen, Fast iterative methods for discrete-ordinates particle transport calculations, Prog. Nucl. Energy 40 (2002) 3-159, https://doi.org/10.1016/S0149-1970(01)00023-3. 
  17. R. Macfarlane, D.W. Muir, R.M. Boicourt, I.A.C. Kahler, J.L. Conlin, The NJOY Nuclear Data Processing System, 2017, https://doi.org/10.2172/1338791.Version 2016, Los Alamos, NM (United States). 
  18. R.E. Macfarlane, TRANSX 2: a code for interfacing MATXS cross-section libraries to nuclear transport codes, Los Alamos Natl. Lab. (1993). LA-12312-MS. 
  19. C.J. Werner, MCNP users manual, Code Version 6.2, Los Alamos Natl. Lab. (2017). LA-UR-29981. 
  20. D.A. Brown, M.B. Chadwick, et al., ENDF/B-VIII.0: the 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data, Nucl. Data Sheets 148 (2018) 1-142, https://doi.org/10.1016/j.nds.2018.02.001. 
  21. D. Wiarda, M.E. Dunn, D.E. Peplow, T.M. Miller, H. Akkurt, Development and testing of ENDF/B-VI. 8 and ENDF/B-VII. 0 coupled neutron-gamma libraries for scale 6, Oak Ridge Natl. Lab. (2009). ORNL/TM-2008/047. 
  22. J.E. White, D.T. Ingersoll, C.O. Slater, R.W. Roussin, BUGLE-96: a revised multigroup cross section library for lwr applications based on ENDF/B-VI release 3, Oak Ridge Natl. Lab. (1996). CONF-960415-37. 
  23. E.E. Lewis, M.A. Smith, N. Tsoulfanidis, G. Palmiotti, T.A. Taiwo, R.N. Blomquist, Benchmark specification for Deterministic 2-D/3-D MOX fuel assembly transport calculations without spatial homogenization (C5G7 MOX), Nucl. Energy Agency, Organisation Econ. Co-operation Dev. 2001, NEA/NSC/DOC (2001) 4. 
  24. K. Kobayashi, N. Sugimura, Y. Nagaya, 3D radiation transport benchmark problems and results for simple geometries with void region, Prog. Nucl. Energy 39 (2001) 119-144, https://doi.org/10.1016/S0149-1970(01)00007-5.