Nuclear Engineering and Technology

  • 제56권3호
  • /
  • Pages.772-784
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  • 2024
  • /
  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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Nodal method for handling irregularly deformed geometries in hexagonal lattice cores

  • Seongchan Kim (School of Mechanical Engineering, Pusan National University) ;
  • Han Gyu Joo (Department of Nuclear Engineering, Seoul National University) ;
  • Hyun Chul Lee (School of Mechanical Engineering, Pusan National University)
  • 투고 : 2023.05.09
  • 심사 : 2023.07.18
  • 발행 : 2024.03.25
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초록

The hexagonal nodal code RENUS has been enhanced to handle irregularly deformed hexagonal assemblies. The underlying RENUS methods involving triangle-based polynomial expansion nodal (T-PEN) and corner point balance (CPB) were extended in a way to use line and surface integrals of polynomials in a deformed hexagonal geometry. The nodal calculation is accelerated by the coarse mesh finite difference (CMFD) formulation extended to unstructured geometry. The accuracy of the unstructured nodal solution was evaluated for a group of 2D SFR core problems in which the assembly corner points are arbitrarily displaced. The RENUS results for the change in nuclear characteristics resulting from fuel deformation were compared with those of the reference McCARD Monte Carlo code. It turned out that the two solutions agree within 18 pcm in reactivity change and 0.46% in assembly power distribution change. These results demonstrate that the proposed unstructured nodal method can accurately model heterogeneous thermal expansion in hexagonal fueled cores.

키워드

과제정보

This work was supported by the National Research Foundation of Korea (NRF) grant funded by the Korea government (MSIT) (No. 2019R1A2C2089962).

참고문헌

  1. L. Giorgio, M. Mauro, T. Nicola, Generation IV nuclear reactors: current status and future prospects, Energy Pol. 61 (2013) 1503-1520.  https://doi.org/10.1016/j.enpol.2013.06.101
  2. G.M. Greenman, Calculation of the Reactivity Feedback Due to Core Assembly Bowing in LMFBRs, Argonne Natl. Lab, IL,USA, 1984. 
  3. P.J. Finck, A Technique for Computing the Reactivity Feedback Due to Core-Assembly Bowing in LMFBR's, Argonne National Lab., No. FRA-TM-159, IL, USA, 1987. 
  4. G. Palmiotti, J.M. Rieunier, C. Gho, M. Salvatores, BISTRO optimized two dimensional Sn transport code, in: Topical Meeting on Advances in Reactor Physics, Mathematics and Computation, 1987. Paris, France, April. 
  5. C. Fiorina, I.D. Clifford, M. Aufiero, K. Mikityuk, GeN-Foam: a novel OpenFOAM based multi-physics solver for 2D/3D transient analysis of nuclear reactors, Nucl. Eng. Des. 294 (2015) 24-37.  https://doi.org/10.1016/j.nucengdes.2015.05.035
  6. E.R. Shemon, et al., NEAMS neutronics: development and validation status, in: Proceedings of ICAPP, 2014. Charlotte, USA, April 6-9 (2014). 
  7. J.I. Yoon, H.G. Joo, Two-level coarse mesh finite difference formulation with multigroup source expansion nodal kernels, J. Nucl. Sci. Technol. 45 (7) (2008) 668-682.  https://doi.org/10.1080/18811248.2008.9711467
  8. J.Y. Cho, et al., Non-linear Triangle-Based Polynomial Expansion Nodal Method for Hexagonal Core Analysis," No. KAERI/TR-1652/2000, Korea Atomic Energy Research Institute, 2000. 
  9. K.S. Kim, M.D. DeHart, Unstructured partial-and net-current based coarse mesh finite difference acceleration applied to the extended step characteristics method in NEWT, Ann. Nucl. Energy 38 (2-3) (2011) 527-534.  https://doi.org/10.1016/j.anucene.2010.09.011
  10. J.J. Grundzinski, C. Grandy, Design and Analysis of the Core Restraint System for a Small Modular Fast Reactor, 2013. Transactions SMiRT-22, San Francisco, CA, USA, August 18-23. 
  11. E.R. Shemon, et al., Direct neutronics modeling approach for deformed core analysis using PROTEUS, Proceedings of M&C (2017). Jeju, Korea, April 16-20 (2017). 
  12. E. Lum, C.L. Pope, Simulation of the fast reactor fuel assembly duct-bowing reactivity effect using Monte Carlo neutron transport and finite element analysis, Nucl. Technol. 207 (5) (2021) 761-770.  https://doi.org/10.1080/00295450.2020.1794190
  13. G. Buckel, K. Kufner, B. Stehle, Benchmark calculations for a sodium-cooled breeder reactor by two and three-dimensional diffusion methods, Nucl. Sci. Eng. 64 (1977) 75-89. https://doi.org/10.13182/NSE77-A27079