Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

A validation study of the SLTHEN code for hexagonal assemblies of wire-wrapped pins using liquid metal heating experiments

  • Sun Rock Choi (Korea Atomic Energy Research Institute) ;
  • Junkyu Han (Korea Atomic Energy Research Institute) ;
  • Huee-Youl Ye (Korea Atomic Energy Research Institute) ;
  • Jonggan Hong (Korea Atomic Energy Research Institute) ;
  • Won Sik Yang (Department of Nuclear Engineering and Radiological Sciences, University of Michigan)
  • Received : 2023.06.07
  • Accepted : 2023.11.08
  • Published : 2024.04.25
Download PDF
⟨ Previous Next ⟩

Abstract

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19-and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

Keywords

Acknowledgement

This work was supported by National Research Foundation of Korea (NRF) grants funded by the Korean government (MSIT) [grant numbers 2021M2E2A1037871and 2021M2E2A2081061].

References

  1. J. Yoo, et al., Overall system description and safety characteristics of prototype Gen IV sodium cooled fast reactor in Korea, Nucl. Eng. Technol. 48 (2016) 1059-1070. https://doi.org/10.1016/j.net.2016.08.004
  2. E.P. Coomes, et al., User Manual COBRA-WC: A Version of COBRA for Single-phase, Single and Multiassembly Thermal-Hydraulic Transient Analysis, PNL-4303, Pacific Northwest Laboratory, Richland, Washington, 1982.
  3. J.N. Lillington, SABRE-3 - A Computer Program for the Calculation of Steady State Boiling in Rod-Clusters, AEEW-M-1647, Unit Kingdom Atomic Energy Authority, 1979.
  4. W.S. Kim, Y.G. Kim, Y.J. Kim, A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly, Ann. Nucl. Energy 29 (2002) 303-321. https://doi.org/10.1016/S0306-4549(01)00041-X
  5. W.S. Yang, An LMR core thermal-hydraulics code based on the ENERGY model, J. Korean Nucl. Soc. 29 (1997) 406-416.
  6. W.S. Yang, Research on the Development of a Computational Code for Steady-State Thermal-Hydraulics Analysis in an LMR Core, vols. 95-83, Basic Electric Power Engineering Research Institute, 1996.
  7. E.U. Khan, W.M. Rohsenow, A.A. Sonin, N.E. Todreas, A porous body model for predicting temperature distribution in wire-wrapped rod assemblies, Nucl. Eng. Des. 35 (1975) 1-12. https://doi.org/10.1016/0029-5493(75)90076-X
  8. E.H. Novendstern, Turbulent flow pressure drop model for fuel rod assemblies utilizing a helical wire-wrap spacer system, Nucl. Eng. Des. 22 (1972) 19-27. https://doi.org/10.1016/0029-5493(72)90059-3
  9. C. Chiu, N.E. Todreas, W.M. Rohsenow, Turbulent flow split model and supporting experiments for wire-wrapped core assemblies, Nucl. Technol. 50 (1990) 40-52. https://doi.org/10.13182/NT80-A17068
  10. S.K. Cheng, N.E. Todreas, Hydrodynamic models and correlations for bare and wire-wrapped hexagonal rod bundles - bundle friction factors, subchannel friction factors and mixing parameters, Nucl. Eng. Des. 92 (1986) 227-251. https://doi.org/10.1016/0029-5493(86)90249-9
  11. S.K. Chang, et al., Flow distribution and pressure loss in subchannels of a wire-wrapped 37-pin rod bundle for a sodium-cooled fast reactor, Nucl. Eng. Technol. 48 (2016) 376-385. https://doi.org/10.1016/j.net.2015.12.013
  12. S.K. Chang, et al., Experimental study of the flow characteristics in an SFR type 61-pin rod bundle using iso-kinetic sampling method, Ann. Nucl. Energy 106 (2017) 160-169. https://doi.org/10.1016/j.anucene.2017.03.024
  13. H. Kim, et al., Flow Mixing Characteristics in Subchannels of a Wire-Wrapped 61-pin Rod Assembly for a Sodium-Cooled Fast Reactor, 16th International Meeting on Nuclear Reactor Thermal Hydraulics, 2015.
  14. H. Kim, et al., Investigations of single-phase flow mixing characteristics in a wire-wrapped 37-pin bundle for a sodium-cooled fast reactor, Ann. Nucl. Energy 87 (2016) 541-546. https://doi.org/10.1016/j.anucene.2015.09.022
  15. S.R. Choi, et al., Assessment of subchannel flow mixing coefficients for wire-wrapped hexagonal fuel rod bundles, Ann. Nucl. Energy 166 (2022), 108810.
  16. B. Chen, N.E. Todreas, Prediction of the coolant temperature field in a breeder reactor including interassembly heat transfer, Nucl. Eng. Des. 35 (1975) 423-440. https://doi.org/10.1016/0029-5493(75)90072-2
  17. K.L. Basehore, N.E. Todreas, SUPERENERGY-2: A Multiassembly, Steady-State Computer Code for LMFBR Core Thermal-Hydraulic Analysis, PNL-3379, Pacific Northwest Laboratory, Richland, Washington, 1982.
  18. W.S. Yang, A.M. Yacout, Assessment of the SE2-ANL Code Using EBR-II Temperatures Measurement, 7th International Meeting on Nuclear Reactor Thermal Hydraulics, 1995.
  19. R.E. Fontana, et al., Temperature distribution in the duct wall and at the exit of a 19-rod simulated LMFBR fuel assembly (FFM bundle 2A), Nucl. Technol. 24 (1974) 176-200. https://doi.org/10.13182/NT74-A31474
  20. R.H. Morris, et al., Single-phase Sodium Tests in 61-pin Full-Length Simulated LMFBR Assembly- Record of Phase 1 Experimental Data for THOR Bundle 9, 1980. ORNL/TM-7313.
  21. F.C. Engel, et al., Characterization of heat transfer and temperature distributions in an electrically heated model of an LMFBR blanket assembly, Nucl. Eng. Des. 62 (1980) 335-347. https://doi.org/10.1016/0029-5493(80)90037-0
  22. J. Pacio, M. Daubner, F. Fellmoser, K. Litfin, Th Wetzel, Experimental study of heavy-liquid metal (LBE) flow and heat transfer along a hexagonal 19-rod bundle with wire spacers, Nucl. Eng. Des. 301 (2016) 111-127. https://doi.org/10.1016/j.nucengdes.2016.03.003
  23. H.A. Abderrahim, P. Baeten, D.D. Bruyn, R. Fernandez, Myrrha - a multi-purpose fast spectrum research reactor, Energy Convers. Manage. 63 (2012) 4-10. https://doi.org/10.1016/j.enconman.2012.02.025
  24. J. Pacio, T. Wetzel, H. Doolaard, F. Roelofs, K. Van Tichelen, Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: experiments and simulations, Nucl. Eng. Des. 312 (2017) 327-337. https://doi.org/10.1016/j.nucengdes.2016.08.023
  25. M.D. Carelli, A.J. Friedland, Hot channel factors for rod temperature calculations in LMFBR assemblies, Nucl. Eng. Des. 62 (1980) 155-180. https://doi.org/10.1016/0029-5493(80)90027-8