Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Application of TULIP/STREAM code in 2-D fast reactor core high-fidelity neutronic analysis

  • Du, Xianan (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Choe, Jiwon (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Choi, Sooyoung (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Lee, Woonghee (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Cherezov, Alexey (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST)) ;
  • Lim, Jaeyong (Korea Atomic Energy Research Institute) ;
  • Lee, Minjae (Korea Atomic Energy Research Institute) ;
  • Lee, Deokjung (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST))
  • Received : 2019.02.28
  • Accepted : 2019.06.08
  • Published : 2019.12.25
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Abstract

The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power distributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic PN scattering, and the multi-group cross-section generation model was performed on the keff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a keff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%.

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References

  1. N.Z. Cho, G.S. Lee, C.J. Park, Fusion of method of characteristics and nodal method for 3-d whole core transport calculation, Trans. Am. Nucl. Soc. 86 (2002) 322-324.
  2. J.Y. Cho, J.S. Song, K.H. Lee, Three dimensional nuclear analysis system DeCART/CHORUS/MASTER, in: ANS Annual Meeting, Atlanta, USA, June 16-20, 2013.
  3. Y.S. Jung, C.B. Shin, C.H. Lim, H.G. Joo, Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers, Ann. Nucl. Energy 62 (2013) 357-374. https://doi.org/10.1016/j.anucene.2013.06.031
  4. B. Collins, S. Stimpson, B.W. Kelley, M.T. Young, B. Kochunas, A. Graham, E.W. Larsen, T. Downar, A. Godfrey, Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT, J. Comput. Phys. 326 (2016) 612-628. https://doi.org/10.1016/j.jcp.2016.08.022
  5. J. Chen, Z.Y. Liu, C. Zhao, Q.M. He, T.J. Zu, L.Z. Cao, H.C. Wu, A new high-fidelity neutronics code NECP-X, Ann. Nucl. Energy 116 (2018) 417-428. https://doi.org/10.1016/j.anucene.2018.02.049
  6. S. Choi, C. Lee, D. Lee, Resonance treatment using pin-based pointwise energy slowing-down method, J. Comput. Phys. 330 (2017) 134-155. https://doi.org/10.1016/j.jcp.2016.11.007
  7. S. Santandrea, L. Graziano, D. Sciannandrone, Accelerated polynomial axial expansions for full 3D neutron transport MOC in the APOLLO3(R) code system as applied to the ASTRID fast breeder reactor, Ann. Nucl. Energy 113 (2018) 194-236. https://doi.org/10.1016/j.anucene.2017.11.010
  8. C.H. Lim, H.G. Joo, W.S. Yang WS, Development of a fast reactor multigroup cross section generation code EXUS-F capable of direct processing of evaluated nuclear data files, Nucl Eng Technol 50 (2018) 340-355, v3. https://doi.org/10.1016/j.net.2018.01.013
  9. C.H. Lee, W.S. Yang, MC2-3: multigroup cross section generation code for fast reactor analysis, Nucl. Sci. Eng. 187 (2017) 268-290. https://doi.org/10.1080/00295639.2017.1320893
  10. W.S. Yang, M.A. Smith, C.H. Lee, A. Wollaber, D. Kaushik, A.S. Mohamed, Neutronics modeling and simulation of SHARP for fast reactor analysis, Nucl Eng Technol 42 (v5) (2010) 520-545. https://doi.org/10.5516/NET.2010.42.5.520
  11. S. Choi S, D. Lee. Efficient parallelization strategy of STREAM for three-dimensional whole-core neutron transport calculation, in: Transactions of the Korean Nuclear Society Spring Meeting, Jeju, Korea, May 17-18, 2018.
  12. K. Kim, S. Choi, D. Lee. Validation of ultra-fine group library generation of lead-cooled fast reactor for STREAM code. In Transactions of the Korean Nuclear Society Spring Meeting, Jeju, Korea, May 17-18, 2018.
  13. X.N. Du, L.Z. Cao, Y.Q. Zheng, H.C. Wu, A hybrid method to generate few-group cross sections for fast reactor analysis, J. Nucl. Sci. Technol. 55 (v8) (2018) 931-944. https://doi.org/10.1080/00223131.2018.1452650
  14. Y.Q. Zheng, X.N. Du, Z.T. Xu, S.C. Zhou, Y. Liu, C.H. Wan, L.F. Xu, SARAX: a new code for fast reactor analysis part I: Methods, Nucl. Eng. Des. 340 (2018) 421-430. https://doi.org/10.1016/j.nucengdes.2018.10.008
  15. Y.Q. Zheng, L. Qiao, Z.A. Zhai, X.N. Du, Z.T. Xu, SARAX: a new code for fast reactor analysis part II: verification, validation and uncertainty quantification, Nucl. Eng. Des. 331 (2018) 41-53. https://doi.org/10.1016/j.nucengdes.2018.02.033
  16. C. Steven, V.D. Marck, Benchmarking ENDF/B-VII.0, Nucl. Data Sheets 107 (2006) 3061-3118. https://doi.org/10.1016/j.nds.2006.11.002
  17. T. Tone, A numerical study of heterogeneity effects in fast reactor critical assemblies, J. Nucl. Sci. Technol. 12 (v8) (1975) 467-481. https://doi.org/10.1080/18811248.1975.9733139
  18. R.E. MacFarlane, D.W. Muir, A.C. Kahler, The NJOY Nuclear Data Processing System, Version 2012, Los Alamos National Laboratory, New Mexico, 2012. LA-UR-12-27079.
  19. A. Yamamoto, M. Tabuchi, N. Sugimura, T. Ushio, M. Mori, Derivation of optimum polar angle quadrature set for the method of characteristics based on approximation error for the bickley function, J. Nucl. Sci. Technol. 44 (v2) (2007) 129-136. https://doi.org/10.3327/jnst.44.129
  20. J.Y. Cho, K.S. Kim, H.J. Shim, J.S. Song, C.C. Lee, H.G. Joo, Whole core transport calculation employing hexagonal modular ray tracing and CMFD formulation, J. Nucl. Sci. Technol. 45 (v8) (2008) 740-751. https://doi.org/10.3327/jnst.45.740
  21. J.R. Jang, W.Y. Kim, S.G. Jeong, E. Jeong, J.S. Park, M. Lemaire, H.S. Lee, Y.M. Jo, P. Zhang, D. Lee, Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask, Ann. Nucl. Energy 114 (2018) 495-509. https://doi.org/10.1016/j.anucene.2017.12.054
  22. S.Y. Choi, J.H. Cho, M.H. Bae, J. Lim, D. Puspitarini, J.H. Jeun, H.G. Joo, S. Hwang, PASCAR: long burning small modular reactor based on natural circulation, Nucl. Eng. Des. 241 (v5) (2011) 1486-1499. https://doi.org/10.1016/j.nucengdes.2011.03.005
  23. OECD/NEA, Benchmark for Neutronic Analysis of Sodium-Cooled Fast Reactor Cores with Various Fuel Types and Core Sizes, vol. 9, NEA/NSC/R, 2015.