Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Effect of two way thermal hydraulic-fuel performance coupling on multicycle depletion

  • Awais Zahur (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology) ;
  • Muhammad Rizwan Ali (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology) ;
  • Deokjung Lee (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology)
  • Received : 2023.04.17
  • Accepted : 2023.08.18
  • Published : 2023.12.25
Download PDF
⟨ Previous Next ⟩

Abstract

A Multiphysics coupling framework, MPCORE, has been developed to analyze safety parameters using the best estimate codes. The framework contains neutron kinetics (NK), thermal hydraulics (TH), and fuel performance (FP) codes to analyze fuel burnup, radial power distribution, and coolant temperature (Tbc). Shuffling and rotation capabilities have been verified on the Watts Bar reactor for three cycles. This study focuses on two coupling approaches for TH and FP modules. The one-way coupling approach involves coupling the FP code with the NK code, providing no data to the TH modules but getting Tbc as boundary condition from TH module. The two-way coupling approach exchanges information from FP to TH modules, so that the simplified heat conduction solver of the TH module is not used. The power profile in both approaches does not differ significantly, but there is an impact on coolant and cladding parameters. The one-way coupling approach tends to over-predict the cladding hydrogen concentration (CHC). This research highlights the difference between one-way and two-way coupling on critical boron concentration, Tbc, CHC, oxide surface temperature, and pellet centerline temperature. Overall, MPCORE framework with two-way coupling provides a more accurate and reliable analysis of safety parameters for nuclear reactors.

Keywords

Acknowledgement

This research was supported by the project (L20S089000) by Korea Hydro and Nuclear Power Co. Ltd. This work was partially supported by Korea Institute of Energy Technology Evaluation and Planning (KETEP) grant funded by the Korea government (MOTIE) [RS-2023-00241302].

References

  1. A.T. Godfrey, VERA Core Physics Benchmark Progression Problem Specifications, Oak Ridge National Laboratory, Oak Ridge, 2014. CASL-U-2012-0131-004.
  2. N. Horelik, et al., Benchmark for Evaluation and Validation of Reactor Simulations, MIT Computational Reactor Physics Group, 2017.
  3. T. Downar, D. Lee, Y. Xu, T. Kozlowski, PARCS v2.6 U.S. NRC Core Neutronics Simulator Theory Manual, School of Nuclear Engineering, Purdue University, 2004.
  4. J. Park, et al., RAST-K v2-three-dimensional nodal diffusion code for pressurized water reactor core analysis, Energies 13 (23) (2020).
  5. J. Choe, et al., Verification and validation of STREAM/RAST-K for PWR analysis, Nucl. Eng. Technol. (2018), https://doi.org/10.1016/j.net.2018.10.004.
  6. J. Jang, et al., Analysis of rostov-II benchmark using conventional two-step code systems, Energies 15 (9) (2022), https://doi.org/10.3390/en15093318.
  7. J. Rhodes, K. Smith, D. Lee, CASMO-5 development and applications, in: Presented at the PHYSOR-2006, Canada, Vancouver, BC, 2006. September 10-14.
  8. S. Choi, D. Lee, Three-dimensional method of characteristics/diamond-difference transport analysis method in STREAM for whole-core neutron transport calculation, Comput. Phys. Commun. 260 (2021), https://doi.org/10.1016/j.cpc.2020.107332.
  9. H.J. Park, S.J. Kim, H. Kwon, J.Y. Cho, BEAVRS benchmark analyses by DeCART stand-alone calculations and comparison with DeCART/MATRA multi-physics coupling calculations, Nucl. Eng. Technol. 52 (9) (2020) 1896-1906.
  10. H. Lee, et al., MCS - a Monte Carlo particle transport code for large-scale power reactor analysis, Ann. Nucl. Energy 139 (2020), https://doi.org/10.1016/j.anucene.2019.107276.
  11. M. Garcia, J. Leppanen, V. Sanchez-Espinoza, A Collision-based Domain Decomposition scheme for large-scale depletion with the Serpent 2 Monte Carlo code, Ann. Nucl. Energy 152 (2021), 108026.
  12. P.K. Romano, N.E. Horelik, B.R. Herman, A.G. Nelson, B. Forget, K. Smith, OpenMC: a state-of-the-art Monte Carlo code for research and development, Ann. Nucl. Energy 82 (2015) 90-97.
  13. H.-J. Shim, B.-S. Han, J.-S. Jung, H.-J. Park, C.-H. Kim, McCARD: Monte Carlo code for advanced reactor design and analysis, Nucl. Eng. Technol. 44 (2) (2012) 161-176.
  14. M. Vazquez, H. Tsige-Tamirat, L. Ammirabile, F. Martin-Fuertes, Coupled neutronics thermal-hydraulics analysis using Monte Carlo and sub-channel codes, Nucl. Eng. Des. 250 (2012) 403-411.
  15. D.J. Kelly, A.E. Kelly, B.N. Aviles, A.T. Godfrey, R.K. Salko, B.S. Collins, MC21/CTF and VERA multiphysics solutions to VERA core physics benchmark progression problems 6 and 7, Nucl. Eng. Technol. 49 (6) (2017) 1326-1338.
  16. M. Garcia, et al., A Serpent2-SUBCHANFLOW-TRANSURANUS coupling for pin-by-pin depletion calculations in Light Water Reactors, Ann. Nucl. Energy 139 (2020), 107213 [Online]. Available: https://www.sciencedirect.com/science/article/pii/S0306454919307236?via%3Dihub.
  17. Z. Liu, B. Wang, M. Zhang, X. Zhou, L. Cao, H. Wu, An internal parallel coupling method based on NECP-X and CTF and analysis of the impact of thermal-hydraulic model to the high-fidelity calculations, Ann. Nucl. Energy 146 (2020), 107645 [Online]. Available: https://www.sciencedirect.com/science/article/pii/S0306454920303431.
  18. K.T. Clarno, et al., Coupled neutronics thermal-hydraulic solution of a full-core PWR using VERA-CS, United States, in: Presented at the Conference: PHYSOR 2014, Kyoto, Japan, 2014, pp. 20140928-20141003 [Online]. Available: https://www.osti.gov/biblio/1154814.
  19. Y.S. Jung, C.B. Shim, C.H. Lim, H.G. Joo, Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers, Ann. Nucl. Energy 62 (2013) 357-374 [Online]. Available: https://www.sciencedirect.com/science/article/pii/S0306454913003344#f0090.
  20. G.K. Delipei, P. Rouxelin, A. Abarca, J. Hou, M. Avramova, K. Ivanov, CTF-PARCS core multi-physics computational framework for efficient LWR steady-state, depletion and transient uncertainty quantification, Energies 15 (14) (2022) 5226 [Online]. Available: https://www.mdpi.com/1996-1073/15/14/5226. 1073/15/14/5226
  21. S. Stimpson, et al., Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1-3, Nucl. Eng. Des. 327 (2018) 172-186.
  22. S. Suman, Influence of hydrogen concentration on burst parameters of Zircaloy-4 cladding tube under simulated loss-of-coolant accident, Nucl. Eng. Technol. 52 (9) (2020) 2047-2053.
  23. I. Davis, O. Courty, M. Avramova, A. Motta, High-fidelity multi-physics coupling for determination of hydride distribution in Zr-4 cladding, Ann. Nucl. Energy 110 (2017) 475-485.
  24. M. Mankosa, M. Avramova, Three-dimensional multi-physics modeling of hydrogen and hydride distribution in zirconium alloy cladding, Prog. Nucl. Energy 105 (2018) 294-300.
  25. R.K. Salko, M.N. Avramova, CTF Pre-processor User's Manual, The Pennsylvania State University, 2016.
  26. A. Cherezov, H. Kim, J. Park, D. Lee, Fuel rod analysis programming interface for loosely coupled multiphysics system, in: Presented at the in Proceedings of the American Nuclear Society Annual Meeting, Chicago, IL, USA, 2020, 16-19 November.
  27. K. Geelhood, W. Luscher, P. Raynaud, I. Porter, FRAPCON-4.0: A Computer Code for the Calculation of Steady-State, Thermal-Mechanical Behavior of Oxide Fuel Rods for High Burnup, vol. 1, Pacific Northwest National Laboratory, PNNL, 2015, 19418.
  28. K. Geelhood, W. Luscher, J. Cuta, I. Porter, in: FRAPTRAN-2.0: A Computer Code for the Transient Analysis of Oxide Fuel Rods vol. 1, Pacific Northwest National Laboratory, PNNL, 2016, 19400.
  29. A. Cherezov, J. Park, H. Kim, J. Choe, D. Lee, A multi-physics adaptive time step coupling algorithm for light-water reactor core transient and accident simulation, Energies 13 (23) (2020), https://doi.org/10.3390/en13236374.
  30. S. Stimpson, et al., Coupled fuel performance calculations in VERA and demonstration on Watts Bar unit 1, cycle 1, Ann. Nucl. Energy 145 (2020), 107554 [Online]. Available: https://www.sciencedirect.com/science/article/pii/S0306454920302528.
  31. J. Yu, H. Lee, M. Lemaire, H. Kim, P. Zhang, D. Lee, MCS based neutronics/thermal-hydraulics/fuel-performance coupling with CTF and FRAPCON, Comput. Phys. Commun. 238 (2019) 1-18.
  32. X. Xu, Z. Liu, H. Wu, L. Cao, Neutronics/thermal-hydraulics/fuel-performance coupling for light water reactors and its application to accident tolerant fuel, Ann. Nucl. Energy 166 (2022), 108809 [Online]. Available: https://www.sciencedirect.com/science/article/pii/S0306454921006861. 1006861
  33. M.I. Radaideh, T. Kozlowski, Combining simulations and data with deep learning and uncertainty quantification for advanced energy modeling, Int. J. Energy Res. 43 (14) (2019) 7866-7890, https://doi.org/10.1002/er.4698.
  34. 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), https://doi.org/10.1016/j.anucene.2019.107271.
  35. B. Roostaii, H. Kazeminejad, S. Khakshournia, Including high burnup structure effect in the UO2 fuel thermal conductivity model, Prog. Nucl. Energy 131 (2021), 103561.
  36. O. Courty, A.T. Motta, J.D. Hales, Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding, J. Nucl. Mater. 452 (1-3) (2014) 311-320 [Online]. Available: https://www.sciencedirect.com/science/article/pii/S0022311514002803.