Authors & Affiliations
Bass L.P.1, Nikolaeva O.V.1, Davidenko V.D.2, Gaifulin S.A.1, Danilov A.A.3, Khmylev A.N.2
1 Keldysh Institute of Applied Mathematics Russian Academy of Sciences, Moscow, Russia
2 National Research Centre “Kurchatov Institute”, Moscow, Russia
3 G.I. Marchuk Institute of Numerical Mathematics of the Russian Academy of Sciences, Moscow, Russia
Nikolaeva O.V.1 — Senior Researcher, Cand. Sci. (Phys.-Math.).
Gaifulin S.A.1 — Researcher.
Davidenko V.D.2 — Director of Department, Dr. Sci. (Techn.).
Khmylev A.N.2 — Leading Researcher, Cand. Sci. (Phys.-Math.
Danilov A.A.3 — Senior Researcher, Cand. Sci. (Phys.-Math.).
The article discusses the issue of the concept of "new generation code", which has been actively used recently to characterize computer programs designed to solve problems of the transfer of neutrons and gamma quanta in nuclear facilities. As an example of a new generation code developed for solving the multigroup transport equation by the grid (deterministic) method, the first version of the new software package RADUGA-TV is considered, including, in particular, the UNK complex for calculating burnup. The article lists the main features of the RADUGA-TV code: the problems to be solved, the types of constants used, the methods for specifying the geometry of the calculation area, the methods for constructing an unstructured spatial mesh. The possibilities of the postprocessor for processing the obtained solution are presented. The article presents progressive algorithms included in the RADUGA-TV code, including grid schemes and methods for parallelizing computations. The advantages of using unstructured grids, including those consisting of cells of various types, are discussed. Methods for parllelizing computations on hybrid computing systems are considered. The question of the spatial grid decomposition when parallelizing computations on distributed memory systems is considered, as well as the question of organizing parallel computation on such systems. Comparison of the characteristics and capabilities of the RADUGA-TV code and other similar in purpose codes, foreign (ATTILA, AE-TIUS, ARES, THOR) and domestic ODETTA is performed. It is shown that the RADUGA-TV code is significantly advanced methodically and practically has no analogues. The article was written based on the materials of the report at the conference "Neutronika-19" and contains more detailed information on the issues discussed in the report.
multi-group transport equation, next generation code, unstructured grids, parallel computation, grid decomposition
1. Bass L.P., Nikolaeva O.V., Davidenko V.V., Gaifulin S.A., Danilov A.A., Khmylev A.N. RADUGA-TV — kod novogo pokoleniya dlya resheniya uravneniya perenosa [RADUGA-TV — a new generation code for solving the transport equation]. Trudy nauchno-tekhnicheskoy konferentsii Neytronno-fizi-cheskiye problemy atomnoy energetiki [Proc. Sci. and Techn. Conf. Neutron-physical Problems of Nuclear Power]. Obninsk, 2019, pp. 84.
2. Belousov V.I., Grushin N.A., Sychugova E.P., Seleznev E.F. Nekotoryye rezul'taty verifikatsii koda Odetta dlya neodnorodnykh zadach [Some results of verification of the Odette code for inhomogeneous problems]. Voprosy atomnoy nauki i tekhniki. Seriya: Fizika yadernykh reaktorov — Problems of Atomic Science and Technology. Series: Physics of Nuclear Reactors, 2018, no. 3, pp. 46–53.
3. Munk M., Slaybaugh R.N. Review of Hybrid Methods for Deep-Penetration Neutron Transport. Nuclear Science and Engineering, 2019, vol. 193, no. 10, pp. 1055–1089. doi.org/10.1080/00295639.2019.1586273.
4. Alcouffe R.E. THREEDANT: A Code to Perform Three-Dimensional, Neutral Particle Transport Calculations. Los Alamos National Laboratory, 1994.
Suslov I.R., Lyamtsev I.A. Gibridnyy metod rascheta zashchity YAEU [Hybrid method for calculating the protection of nuclear power plants]. Preprint FEI-3267 — Preprint IIPE-3267. Obninsk, 2016. 20 p.
6. Wooten D., Powers J.J. A Review of Molten Salt Reactor Kinetics Models. Nuclear Science and Engineering, 2018, vol. 191, no. 3, pp. 203–230. doi.org/10.1080/00295639.2018.1480182.
Kramarenko V.K. Metody resheniya uravneniya diffuzii v sredakh s kontrastnymi vklyucheniyami i s uchetom osobennostey ot raspredelennykh istochnikov. Diss. k.f-m.n. [Methods for solving the diffusion equation in media with contrasting inclusions and taking into account the features from distributed sources. Cand. phys. and math. sci. diss.]. Moscow, 2019. 94 p.
8. Nikolaeva O.V., Kazantseva A.N. Tochnost' skhem metoda konechnykh elementov dlya resheniya uravneniya perenosa na nestrukturirovannykh tetraedricheskikh i prizmaticheskikh setkakh [Accuracy of Finite Element Method Schemes for Solving the Transport Equation on Unstructured Tetrahedral and Prismatic Grids]. Voprosy atomnoy nauki i tekhniki. Seriya: Matematicheskoye modelirovaniye fizicheskikh protsessov – Problems of Atomic Science and Technology. Series: Mathematical modeling of physical processes, 2020, no. 1, pp. 3–19.
9. Nikolaeva O.V., Gaifulin S.A., Bass L.P. O dekompozitsii nestrukturirovannoy setki pri reshenii uravneniya perenosa neytronov na parallel'nykh komp'yuterakh. Parallel'nyye vychislitel'nyye tekhnologii [On the decomposition of an unstructured grid when solving the neutron transport equation on parallel computers]. Trudy mezhdunarodnoy nauchnoy konferentsii “Parallel'nyye vychislitel'nyye tekhnologii (PaVT’2019)” [Proc. Int. Sci. Conf. “Parallel Computational Technologies (PaVT'2019)”]. Kaliningrad, 2019, pp. 362–372.
10. Belousov N.I., Davidenko V.D., Tsibulsky V.F. Programma UNK dlya detal'nogo rascheta spektra neytronov v yacheyke yadernogo reaktora [UNK program for detailed calculation of the neutron spectrum in a nuclear reactor cell]. Preprint IAE-6083/4. Moscow, 1998.
11. Takeda T., Ikeda H. 3-D Neutron Transport Benchmarks. Journal of Nuclear Science and Technology, 1991, vol. 8, no. 7, pp. 656–669. doi.org/10.1080/18811248.1991.9731408.
12. Wareing T.A., McGhee J.M., Morel J.E., ATTILA: A Three–Dimensional Unstructured Tetrahedral-Mesh Discrete Ordinates Code. Transactions of the American Nuclear Society. 1996. Vol. 75. Pp. 146–147.
13. Wareing T.A., McGhee J.M., Morel J.E., Pautz S.P. Discontinuous Finite Element SN Methods on Three-Dimensional Unstructured Grids. Nuclear Science and Engineering, 2001, vol. 138, pp. 256–268. doi.org/10.13182/NSE138-256.
14. Lucas D.S., Gougar H.D., Wareing T., Failla G., McGhee J., Barnett D.A., Davis I. Comparison of the 3-D Deterministic Neutron Transport Code Attila to Measure Data, MCNP and MCNPX for the Advanced Test Reactor. Proc. M&C 2005 International Topical. Avignon, France, 2005.
15. Lucas D.S., Gougar H.D., Roth P.A., Wareing T., Failla G., McGhee J., Barnett A. Applications of the 3-D Deterministic Transport Attila for Core Safety Analysis. Proc. Americas Nuclear Energy Symposium. Miami Beach, Florida, 2004. Available at: https://www.researchgate.net/publica-tion/237614641_Applications_Of_The_3-D_Deterministic_Transport_AttilaR_For_Core_Safety_Anal-ysis/link/02e7e5267dd5feb36c000000/download (accessed 19.09.2020).
16. Vassiliev O.D., Wareing T.A., Davis I.M., McGhee J., Barnett D., Horton J.L., Gifford K., Failla G., Titt U., Mourtada F. Feasibility of a Multigroup Deterministic Solution Method for 3D Radiotherapy Dose Calculations. International Journal of Radiative Oncology, Biology, Physics, 2008, vol. 72, pp. 220–227. doi.org/10.1016/j.ijrobp.2008.04.0572017.
17. Kawrakow I. Accurate condensed history Monte Carlo simulation of electron transport. I. EGSnrc, +the new EGS4 version. Medical Physics, 2000, vol. 27, pp. 485–498. doi.org/10.1118/1.598917.
18. Kim J.W., Lee Y.O. A Deep Penetration Problem Calculation Using AETIUS: An Easy Modeling Discrete Ordinates Transport Code UsIng Unstructured Tetrahedral Mesh, Shared Memory Parallel. EPJ Web of Conferences, 2017, vol. 153, pp. 06025. doi.org/10.1051/epjconf/20171530.
19. Chen Y., Zhang B., Zhang L., Zheng J., Zheng Y., Liu C. ARES: A Parallel Discrete Ordinates Transport Code for Radiation Shielding Applications and Reactor Physics Analysis. Hindawi Science and Technology of Nuclear Installations, 2017, article ID 2596727. 11 p. doi.org/10.1155/2017/2596727.
20. Ferrer R.M., Azmy Y.Y. A Robust Arbitrarily High-Order Transport Method of the Characteristic Type for Unstructured Grids. Nuclear Science and Engineering, 2012, vol. 172, pp. 33–51. doi.org/10.13182/NSE10-106.
21. Yessayan R., Azmy Y., Schunert S. Development of a Parallel Performance Model for the THOR Neutral Particle Transport Code. Mathematics & Computational Methods Applied to Nuclear Science & Engineering: Proc. of the International Conference. Jeju, Korea, 2017. Available at: https://www.osti.gov/servlets/purl/1369430 (accessed 19.09.2020).