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http://dx.doi.org/10.1016/j.net.2021.03.005

Use of Monte Carlo code MCS for multigroup cross section generation for fast reactor analysis  

Nguyen, Tung Dong Cao (Department of Nuclear Engineering Ulsan National Institute of Science and Technology (UNIST))
Lee, Hyunsuk (Department of Nuclear Engineering Ulsan National Institute of Science and Technology (UNIST))
Lee, Deokjung (Department of Nuclear Engineering Ulsan National Institute of Science and Technology (UNIST))
Publication Information
Nuclear Engineering and Technology / v.53, no.9, 2021 , pp. 2788-2802 More about this Journal
Abstract
Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MWth metal-fueled SFR (MET-1000) and a 3600 MWth oxide-fueled SFR (MOX-3600). The accuracy of a few-group XSs generated by MCS is verified using another MC code, Serpent 2. The neutronic steady-state whole-core problem is analyzed using MCS/RAST-K with a 24-group XS set. Various core parameters of interest (core keff, power profiles, and reactivity feedback coefficients) are obtained using both MCS/RAST-K and MCS. A code-to-code comparison indicates excellent agreement between the nodal diffusion solution and stochastic solution; the error in the core keff is less than 110 pcm, the root-mean-square error of the power profiles is within 1.0%, and the error of the reactivity feedback coefficients is within three standard deviations. Furthermore, using the super-homogenization-corrected XSs improves the prediction accuracy of the control rod worth and power profiles with all rods in. Therefore, the results demonstrate that employing the MCS MG XSs for the nodal diffusion code is feasible for high-fidelity analyses of fast reactors.
Keywords
Monte Carlo; Nodal diffusion; Multigroup; Cross section generation; SFR; MCS;
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1 J.E. Kelly, Generation IV International Forum: a decade of progress through international cooperation, Prog. Nucl. Energy 77 (2014) 240-246.   DOI
2 E. Nikitin, E. Fridman, K. Mikityuk, Solution of the OECD/NEA neutronic SFR benchmark with Serpent-DYN3D and Serpent-PARCS code systems, Ann. Nucl. Energy 75 (2015) 492-497.   DOI
3 E. Nikitin, E. Fridman, Extension of the reactor dynamics code DYN3D to SFR applications-Part II: validation against the Phenix EOL control rod withdrawal tests, Ann. Nucl. Energy 119 (2018) 411-418.   DOI
4 W. Heo, W. Kim, Y. Kim, S. Yun, Feasibility of a Monte Carlo-deterministic hybrid method for fast reactor analysis, in: Proceedings of the 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering - M&C 2013, Idaho, USA, May 5-9, 2013.
5 R.S. Sen, A.J. Hummel, H. Hiruta, SuPer-Homogenization (SPH) Corrected Cross Section Generation For High Temperature Reactor (No. INL/EXT-17-41516), Idaho National Laboratory, Idaho Falls, ID (United States), 2017.
6 H. Lee, W. Kim, P. Zhang, M. Lemaire, A. Khassenov, J. Yu, Y. Jo, J. Park, D. Lee, MCS-A Monte Carlo particle transport code for large-scale power reactor analysis, Ann. Nucl. Energy 139 (2020), 107276.   DOI
7 E. Nikitin, E. Fridman, K. Mikityuk, On the use of the SPH method in nodal diffusion analyses of SFR cores, Ann. Nucl. Energy 85 (2015) 544-551.   DOI
8 J. Jang, W. Kim, S. Jeong, E. Jeong, J. Park, M. Lemaire, H. Lee, Y. 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.   DOI
9 L. Ghasabyan, Use of Serpent Monte-Carlo Code for Development of 3D Full-Core Models of Gen-IV Fast-Spectrum Reactors and Preparation of Group Constants for Transient Analyses with PARCS/TRACE Coupled System, master of science thesis, Royal Institute of Technology, KTH, 2013.
10 T.D.C. Nguyen, H. Lee, S. Choi, D. Lee, Validation of UNIST Monte Carlo code MCS using VERA progression problems, Nucl. Eng. Technol. 52 (5) (2020) 878-888.   DOI
11 T.D.C. Nguyen, H. Lee, S. Choi, D. Lee, MCS/TH1D analysis of VERA whole-core multi-cycle depletion problems, Ann. Nucl. Energy 139 (2020), 107271.   DOI
12 V. Dos, H. Lee, J. Choe, M. Lemaire, H.C. Shin, H.S. Lee, D. Lee, Verification & validation of MCS multi-physics analysis capability for OPR-1000 multi-cycle operation, in: Proceedings of the 2019 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering - M&C 2019, Oregon, USA, August 25-29, 2019.
13 N.E. Stauff, T.K. Kim, T.A. Taiwo, et al., Benchmark For Neutronic Analysis of Sodium-Cooled Fast Reactor Cores With Various Fuel Types and Core Sizes (No. NEA-NSC-R-2015-9), Organization for Economic Co-Operation and Development, 2016.
14 T.D.C. Nguyen, H. Lee, J. Choe, M. Lemaire, D. Lee, APR-1400 whole-core depletion analysis with MCS, in: Proceedings of the 2019 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering - M&C 2019, Oregon, USA, August 25-29, 2019.
15 J. Choe, S. Choi, P. Zhang, J. Park, W. Kim, H.C. Shin, H.S. Lee, J.-E. Jung, D. Lee, Verification and validation of STREAM/RAST-K for PWR analysis, Nucl. Eng. Technol. 51 (2) (2019) 356-368.   DOI
16 T.Q. Tran, A. Cherezov, X. Du, J. Park, D. Lee, Development of hexagonal-Z geometry capability in RAST-K for fast reactor analysis, in: 19th International Conference on Emerging Nuclear Energy Systems (ICENES 2019), Bali, Indonesia, October 6-9, 2019.
17 W.R.D. Boyd III, Reactor agnostic Multi-Group Cross Section Generation for Fine-Mesh Deterministic Neutron Transport Simulations, doctoral dissertation, Massachusetts Institute of Technology, 2017.
18 J. Leppanen, M. Pusa, E. Fridman, Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code, Ann. Nucl. Energy 96 (2016) 126-136.   DOI
19 T.M.N. Nguyen, Y. Jo, H. Lee, A. Cherezov, D. Lee, Whole-core Monte Carlo analysis of MOX-3600 core in NEA-SFR benchmark using MCS code, in: Proceedings of the Korean Nuclear Society Autumn Meeting, Yeosu, Korean, October 25-26, 2018.