과제정보
This work was supported by the National Research Foundation of Korea (NRF) grant funded by the Korea government (MSIT). (No.NRF-2017M2A8A2018595).
참고문헌
- J.E. Kelly, Generation IV International Forum: a decade of progress through international cooperation, Prog. Nucl. Energy 77 (2014) 240-246. https://doi.org/10.1016/j.pnucene.2014.02.010
- 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. https://doi.org/10.1016/j.anucene.2014.08.054
- 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. https://doi.org/10.1016/j.anucene.2018.05.016
- 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.
- 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. https://doi.org/10.1016/j.anucene.2015.06.007
- 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.
- 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. https://doi.org/10.1016/j.anucene.2019.107276
- 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. https://doi.org/10.1016/j.net.2019.10.023
- 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. https://doi.org/10.1016/j.anucene.2017.12.054
- 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.
- 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. https://doi.org/10.1016/j.anucene.2019.107271
- 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.
- 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.
- 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. https://doi.org/10.1016/j.net.2018.10.004
- 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.
- 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. https://doi.org/10.1016/j.anucene.2016.06.007
- 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.
- 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.
- 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.