STRAUM-MATXST: A Code System for Multi-group Neutron-Gamma Coupled Transport Calculation with Unstructured Tetrahedral Meshes
0
조회
0
다운로드
0
인용
0
참고문헌
키워드
초록
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.
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.
참고문헌 (24)
- T. A. Wareing, "ATTILA: a three-dimensional, unstructured tetrahedral mesh discrete ordinates transport code", Trans. Am. Nucl. Soc, 75 (1996)
- R. L. Martz, "The MCNP6 book on unstructured mesh geometry: user's guide", Los Alamos Natl. Lab, pp.1~77 (2014)[DOI]
- T.D. Long, "Implementation and Verification of adjoint neutron transport calculation in MUST code", Transactions of the Korea Nuclear Society Virtual Spring Meeting (2020)
- T. Wareing, "Discontinuous finite element SN methods on three-dimensional unstructured grids", Nucl. Sci. Eng, 138, pp.256~268 (2001)[DOI]
- S. G. Hong, "Two subcell balance methods for solving the multigroup discrete ordinates transport equation with tetrahedral meshes", Nucl. Sci. Eng, 173, pp.13182~ (2013)
- H. Muhammad, "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, pp.145~153 (2019)[DOI]
- Y. Azmy, "Nuclear Computational Science: a Century in Review" (2010), Springer
- R. N. Slaybaugh, "Multigrid in energy preconditioner for Krylov solvers", J. Comput. Phys, 242, pp.405~419 (2013)
- "Christophe Geuzaine, Jean-Francois Remacle, Gmsh"[URL]
- M.H. Woo, "Development of multi-group cross section processing program for MUST unstructured discrete ordinate transport code", Transactions of the Korea Nuclear Society Virtual Autumn Meeting (2020)
- T. W. Huang, "Taskflow: a lightweight parallel and heterogeneous task graph computing system", IEEE Trans. Parallel Distrib. Syst, 33, pp.3104255~ (2022)
- Y. Saad, "GMRES: a generalized minimal residual algorithm for solving nonsymmetric linear systems", SIAM J. Sci. Stat. Comput, 7, pp.1137~ (1986)
- 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, pp.1137~ (1992)
- Habib Muhammad, "Diffusion synthetic acceleration with the fi ne mesh rebalance of the subcell balance method with tetrahedral meshes for S N transport calculations", Nuclear Engineering and Technology, 52(3), pp.485~498 (2020), 한국원자력학회
- "Eigen v3, others" (2010)[URL]
- M. L. Adams, "Fast iterative methods for discrete-ordinates particle transport calculations", Prog. Nucl. Energy, 40, pp.3~159 (2002)[DOI]
- R. Macfarlane, "The NJOY Nuclear Data Processing System", ~ (2017)
- R. E. Macfarlane, "TRANSX 2: a code for interfacing MATXS cross-section libraries to nuclear transport codes", ~ (1993)
- MCNP users manual, Code Version 6.2, "C. J. Werner", Los Alamos Natl. Lab, ~ (2017)
- D. A. Brown, "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, 2018, pp.148, 1~142
- Development and testing of ENDF, "B-VI. 8 and ENDF", B-VII. 0 coupled neutron-gamma libraries for scale 6, Oak Ridge Natl. Lab, pp.2009, ~[DOI]
- BUGLE-96: a revised multigroup cross section library for lwr applications based on ENDF, "B-VI release 3", J. E. White, 1996, CONF-960415-37
- Benchmark specification for Deterministic 2-D, "3-D MOX fuel assembly transport calculations without spatial homogenization (C5G7 MOX)", E.E. Lewis, 2001
- K. Kobayashi, "3D radiation transport benchmark problems and results for simple geometries with void region", Prog. Nucl. Energy, 39, pp.119~144 (2001)[DOI]