Vu Thanh Mai, Donny Hartanto, Tran The Anh, Luu Thi Lan, Tran Viet Phu, Pham Nhu Viet Ha

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In this study, the SCALE/TRITON code (based on deterministic method) and the Serpent 2 code (based on Monte Carlo method) were utilized to prepare the group constants of the pressurized water reactor (PWR) mixed-oxide (MOX) fuel assemblies for transient analyses of PWR MOX fueled cores in normal operation and control rod ejection accident condition with 3D reactor kinetics codes. The PWR MOX fuel assemblies were modeled with TRITON and Serpent, and their infinite neutron multiplication factors (k-inf) versus burnup and respective two-group neutron cross sections were calculated and compared against the available benchmark data obtained with the HELIOS code. The comparative results generally show a good agreement between TRITON and Serpent with HELIOS within 643 pcm for the k-inf values and within 5% for the two-group neutron cross sections. Therefore, it indicates that the TRITON and Serpent models developed herein for the PWR MOX fuel assemblies can be applied to group constant generation to be further used in transient analyses of PWR MOX fueled cores.

Keywords: PWR MOX fuel assembly, group constant, SCALE, Serpent.


[1] T. Kozlowski, T. J. Downar, PWR MOX/UO2 Core Transient Benchmark Final Report, Purdue University, ISBN 92-64-02330-5, NEA/NSC/DOC 20, US, 2006.
[2] T. J. Downar, Y. Xu, V. Seker, N. Hudson, User Manual for the PARCS v3.0 U.S. NRC Core Neutronics Simulator, University of Michigan, UM-NERS-09-0001, US, 2012.
[3] S. Pinem, T. M. Sembiring, P. H. Liem, The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark, Science and Technology of Nuclear Installations, Vol. 2014, 2014, Article ID 845832,
[4] B. T. Rearden , M. A. Jessee (Editors), SCALE Code System User Manual (Version 6.2.2), Oak Ridge National Laboratory, ORNL/TM-2005/39, US, 2017.
[5] J. Leppänen, M. Pusa, T. Viitanen, V. Valtavirta, T. Kaltiaisenaho, The Serpent Monte Carlo Code: Status, Development and Applications in 2013, Annals of Nuclear Energy, Vol. 82, 2015, pp. 142-150,
[6] J. Leppänen, M. Pusa, E. Fridman, Overview of Methodology for Spatial Homogenization in the Serpent 2 Monte Carlo Code, Annals of Nuclear Energy, Vol. 96, 2016, pp. 126-136,
[7] Z. Liu, K. Smith, B. Forget, J. Ortensi, Cumulative Migration Method for Computing Rigorous Diffusion Coefficients and Transport Cross Sections from Monte Carlo, Annals of Nuclear Energy, Vol. 112, 2018, pp. 507-516,
[8] J. Leppänen, Input Syntax Manual,, 2020 (accessed on: August 6th, 2020).
[9] M. B. Chadwick, P. Obložinský, M. Herman, N. M. Greene, R. D. McKnight, D. L. Smith, P. G. Young, R. E. MacFarlane, G. M. Hale, S. C. Frankle, A. C. Kahler, T. Kawano, R. C. Little, D. G. Madland, P. Moller, R. D. Mosteller, P. R. Page, P. Talou, H. Trellue, M. C. White, W. B. Wilson, R. Arcilla, C. L. Dunford, S. F. Mughabghab, B. Pritychenko, D. Rochman, A. A. Sonzogni, C. R. Lubitz, T. H. Trumbull, J. P. Weinman, D. A. Brown, D. E. Cullen, D. P. Heinrichs, D. P. McNabb, H. Derrien, M. E. Dunn, N. M. Larson, L. C. Leal, A. D. Carlson, R. C. Block, J. B. Briggs, E. T. Cheng, H. C. Huria, M. L. Zerkle, K. S. Kozier, A. Courcelle, V. Pronyaev, S. C. van der Marck, ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology, Nuclear Data Sheets, Vol. 107, No. 12, 2006, pp. 2931-3060,
[10] J. Leppänen, WIMS 172-group Structure,, 2020 (accessed on: August 6th, 2020).