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http://dx.doi.org/10.1016/j.net.2018.01.013

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)
Publication Information
Nuclear Engineering and Technology / v.50, no.3, 2018 , pp. 340-355 More about this Journal
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.
Keywords
Advanced Burner Reactor 1000 Benchmark; Evaluated Nuclear Data File Format; Fast Reactors; Multigroup Cross Sections; Ultrafine Group;
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Times Cited By KSCI : 4  (Citation Analysis)
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1 C. Lee, W.S. Yang, MC2-3: multigroup cross section generation code for fast reactor analysis, Nuclear Sci. Eng. 187 (2017) 268-290.   DOI
2 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.
3 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.   DOI
4 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.
5 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.   DOI
6 D.E. Cullen, Program SIGMA1 (Version 74-1), Lawrence Livermore Laboratory report, UCID-16426, 1974.
7 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.
8 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.   DOI
9 K. Shibata, et al., JENDL-4.0: a new library for nuclear science engineering, J. Nuclear Sci. Technol. 48 (1) (2011) 1-30.   DOI
10 A. Konig, et al., The JEFF-3.1 Nuclear Data Library, JEFF Report 21, OECD/NEA, 2006.
11 M. Herman, A. Trkov, ENDF Formats Manual, BNL-90365-2009, Brookhaven National Laboratory, 2011.
12 H.J. Shim, et al., McCARD: Monte Carlo code for advanced reactor design and analysis, Nuclear Eng. Technol. 44 (2012) 161-176.   DOI
13 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.
14 R.J.J. Stamm'ler, M.J. Abbate, Methods of Steady-state Reactor Physics in Nuclear Design, Academic Press, London, 1983.
15 Min Ryu, et al., Incorporation of anisotropic scattering in nTRACER, in: Korea Nuclear Society Autumn Meeting, Pyeongchang, Korea, Oct 29-31, 2014.
16 T.K. Kim, et al., Core design studies for a 1000 MWth advanced burner reactor, Ann. Nuclear Energy (2009).
17 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.   DOI
18 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.   DOI
19 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.
20 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.   DOI
21 W.S. Yang, Fast reactor physics and computational methods, Nuclear Eng. Technol. 44 (2) (2017) 177-198.   DOI
22 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.
23 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.