Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Development of a fast reactor multigroup cross section generation code EXUS-F capable of direct processing of evaluated nuclear data files

  • Lim, Changhyun (Department of Nuclear Engineering, Seoul National University) ;
  • Joo, Han Gyu (Department of Nuclear Engineering, Seoul National University) ;
  • Yang, Won Sik (Department of Nuclear Engineering and Radiological Sciences, University of Michigan)
  • Received : 2017.12.15
  • Accepted : 2018.01.08
  • Published : 2018.04.25
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Abstract

The methods and performance of a fast reactor multigroup cross section (XS) generation code EXUS-F are described that is capable of directly processing Evaluated Nuclear Data File format nuclear data files. RECONR of NJOY is used to generate pointwise XS data, and Doppler broadening is incorporated by the Gauss-Hermite quadrature method. The self-shielding effect is incorporated in the ultrafine group XSs in the resolved and unresolved resonance ranges. Functions to generate scattering transfer matrices and fission spectrum matrices are realized. The extended transport approximation is used in zero-dimensional calculations, whereas the collision probability method and the method of characteristics are used for one-dimensional cylindrical geometry and two-dimensional hexagonal geometry problems, respectively. Verification calculations are performed first for various homogeneous mixtures and cylindrical problems. It is confirmed that the spectrum calculations and the corresponding multigroup XS generations are performed adequately in that the reactivity errors are less than 50 pcm with the McCARD Monte Carlo solutions. The nTRACER core calculations are performed with the EXUS-F-generated 47 group XSs for the two-dimensional Advanced Burner Reactor 1000 benchmark problem. The reactivity error of 160 pcm and the root mean square error of the pin powers of 0.7% indicate that EXUF-F generates properly the broad-group XSs.

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References

  1. J. Yoo, et al., Overall system description and safety characteristics of prototype gen IV sodium cooled fast reactor in Korea, Nuclear Eng. Technol. 48 (2016) 1059-1070. https://doi.org/10.1016/j.net.2016.08.004
  2. Yeon Sang Jung, Cheon Bo Shim, Chang Hyun Lim, Han Gyu Joo, Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers, Ann. Nuclear Energy 62 (2013) 357-374. https://doi.org/10.1016/j.anucene.2013.06.031
  3. W.S. Yang, Fast reactor physics and computational methods, Nuclear Eng. Technol. 44 (2) (2017) 177-198. https://doi.org/10.5516/NET.01.2012.504
  4. C.S. Gil, ZZ KAFAX-E70, 150 and 12 groups cross section library in MATXS format based on ENDF/B-VII.0 for fast reactors, NEA-1815, OECD/NEA Data Bank, Issy-les-Moulineaux, France, 2009.
  5. B.J. Toppel, H. Henryson II, C.G. Stenberg, $ETOE-2/MC^{2}-2/SDX$ multi-group cross-section processing, in: Conf-780334-5, Proc. of RSIC Seminarworkshop on Multi-group Cross Sections, Oak Ridge, TN, March, 1978.
  6. C. Lee, W.S. Yang, MC2-3: multigroup cross section generation code for fast reactor analysis, Nuclear Sci. Eng. 187 (2017) 268-290. https://doi.org/10.1080/00295639.2017.1320893
  7. G. Rimpault, Algorithmic features of the ECCO cell code for treating heterogeneous fast reactor subassemblies, in: International Topical Meeting on Reactor Physics and Computations, Portland, Oregon, May 1-5, 1995.
  8. T. Hazama, et al., Development of a fine and ultra-fine group cell calculation code SLAROM-UF for fast reactor analyses, J. Nuclear Sci. Technol. 43 (8) (2016) 908-918. https://doi.org/10.3327/jnst.43.908
  9. R.E. MacFarlane, D.W. Muir, et al., The NJOY Nuclear Data Processing System, Version 2016, LA-UR-17-20093, Los Alamos National Laboratory, 2016.
  10. C. Dean, R. Perry, R. Neal, A. Kyrieleis, Validation of run-time Doppler broadening in MONK with JEFF3.1, J. Korean Phys. Soc. 59 (2) (2011) 1162-1165. https://doi.org/10.3938/jkps.59.1162
  11. D.E. Cullen, Program SIGMA1 (Version 74-1), Lawrence Livermore Laboratory report, UCID-16426, 1974.
  12. C. Lee, W.S. Yang, An Improved Resonance Self-shielding Method for Heterogeneous Fast Reactor Assembly and Core Calculations, M&C 2013, Sun Valley, Idaho, May 5-9, 2013.
  13. M.B. Chadwick, et al., ENDF/B-VII.0: next generation evaluated nuclear data library for nuclear science and technology, Nuclear Data Sheets 107 (12) (2006) 2931-3060. https://doi.org/10.1016/j.nds.2006.11.001
  14. K. Shibata, et al., JENDL-4.0: a new library for nuclear science engineering, J. Nuclear Sci. Technol. 48 (1) (2011) 1-30. https://doi.org/10.1080/18811248.2011.9711675
  15. A. Konig, et al., The JEFF-3.1 Nuclear Data Library, JEFF Report 21, OECD/NEA, 2006.
  16. M. Herman, A. Trkov, ENDF Formats Manual, BNL-90365-2009, Brookhaven National Laboratory, 2011.
  17. R.E. MacFarlane, TRANSX 2: a Code for Interfacing MATXS Cross-section Libraries to Nuclear Transport Codes, LA-12312-MS, Los Alamos National Laboratory, 1992.
  18. C. Lim, H.G. Joo, W.S. Yang, Applications of the probability table on deterministic code for unresolved resonance self-shielding, in: Transactions of the American Nuclear Society Winter Meeting, Washington, U.S, Oct 29 - Nov 2, 2017.
  19. R.J.J. Stamm'ler, M.J. Abbate, Methods of Steady-state Reactor Physics in Nuclear Design, Academic Press, London, 1983.
  20. Min Ryu, et al., Incorporation of anisotropic scattering in nTRACER, in: Korea Nuclear Society Autumn Meeting, Pyeongchang, Korea, Oct 29-31, 2014.
  21. H.J. Shim, et al., McCARD: Monte Carlo code for advanced reactor design and analysis, Nuclear Eng. Technol. 44 (2012) 161-176. https://doi.org/10.5516/NET.01.2012.503
  22. T.K. Kim, et al., Core design studies for a 1000 MWth advanced burner reactor, Ann. Nuclear Energy (2009).
  23. M.B. Chadwick, et al., ENDF/B-VII.1 nuclear data for science and technology: cross sections, covariances, fission product yields and decay data, Nuclear Data Sheets 112 (12) (2011) 2887-2996. https://doi.org/10.1016/j.nds.2011.11.002

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