Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT) (방사성폐기물학회지)

Korean Radioactive Waste Society (한국방사성폐기물학회)
권호 표지
DOI QR코드

DOI QR Code

Sensitivity Analysis of Thermal Parameters Affecting the Peak Cladding Temperature of Fuel Assembly

  • Received : 2023.04.03
  • Accepted : 2023.06.19
  • Published : 2023.09.30
Download PDF
⟨ Previous Next ⟩

Abstract

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

Keywords

Acknowledgement

This work was supported by the Institute for Korea Spent Nuclear Fuel (iKSNF) and the National Research Foundation of Korea (NRF), grant funded by the Korean government (Ministry of Science and ICT, MSIT) (Grant number 2021M2E1A1085226).

References

  1. International Atomic Energy Agency, Storage of Spent Nuclear Fuel, IAEA Safety Standard Series No. SSG-15 (Rev.1) (2012). 
  2. J.A. Fort, T.E. Michener, S.R. Suffield, and D.J. Richmond. Thermal Modeling of a Loaded MAGNASTOR Storage System at Catawba Nuclear Station, Fuel Cycle Research and Development, U.S. Department of Energy Technical Report, PNNL-25871 Rev. 0 (2016). 
  3. R.A. Brewster, E. Baglietto, E. Volpenhein, and C.S. Bajwa, "CFD Analysis of the TN-24P PWR Spent Fuel Storage Cask", Proc. of the ASME 2012 Pressure Vessels and Piping Division Conference, PVP2012-78491, 17-25, July 15-19, 2012, Toronto. 
  4. J. Kim, D. Kook, J. Shim, and Y. Kim, "Review on Spent Nuclear Fuel Performance and Degradation Mechanism Under Long-term Dry Storage", J. Nucl. Fuel Cycle Waste Technol., 11(4), 333-349 (2013).  https://doi.org/10.7733/jnfcwt-k.2013.11.4.333
  5. B. Hanson, H. Alsaed, C. Stockman, D. Enos, R. Mayer, and K. Sorenson, "Gap Analysis to Support Extended Storage of Used Nuclear Fuel Rev. 0", U.S. Department of Energy, PNNL-20509 (2012). 
  6. T.E. Michener, D.R. Rector, J.M. Cuta, and H.E. Adkins, Jr. COBRA-SFS: A Thermal-Hydraulic Analysis Code for Spent Fuel and Storage and Transportation Casks, Cycle 4a, Pacific Northwest Laboratory Report, 153, PNNL-24841(2017). 
  7. D.S. Rowe, COBRA-IIIC: A Digital Computer Program for Steady State and Transient Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements, Battelle Northwest Laboratory, BNWL-1695 (1973). 
  8. K.R. Robb, J.M. Cuta, and L.P. Miller, "Thermal Analysis Capability of UNF-ST&DARDS", Nucl. Technol., 199(3), 289-298 (2017).  https://doi.org/10.1080/00295450.2017.1346446
  9. N.J. Lombardo, J.M. Cuta, T.E. Michener, D.R. Rector, and C.L. Wheeler. COBRA-SFS: A Thermal-Hydraulic Analysis Computer Code, Vol III: Validation Assessments, Pacific Northwest Laboratory Technical Report, PNL-6049 (1986). 
  10. D.R. Rector, R.A. McCann, U.P. Jenquin, C.M. Heeb, J.M. Creer, and C.L. Wheeler. Castor-1C Spent Fuel Storage Cask Decay Heat, Heat Transfer, and Shielding Analyses, Pacific Northwest Laboratory Technical Report, PNL-5974 (1986). 
  11. D.R. Rector, C.L. Wheeler, and N.J. Lombardo. Thermal-Hydraulic Analysis of Spent Fuel Storage Systems, Pacific Northwest Laboratory Technical Report, PNL-SA-14444 (1987). 
  12. T.E. Michener, D.R. Rector, and J.M. Cuta, "Validation of COBRA-SFS With Measured Temperature Data From Spent Fuel Storage Cask", Nucl. Technol., 199(3), 350-368 (2017).  https://doi.org/10.1080/00295450.2017.1327253
  13. M. Sevecek, M. Valach, and C.H. Lee, "Thermal Analysis of Dry Storage and Transportation Casks CASTOR Using COBRA-SFS", Acta Polytechnica CTU Proceedings, vol. 28, 32-41, Czech Technical University, Prague (2020).  https://doi.org/10.14311/APP.2020.28.0032
  14. D.R. Rector. RADGEN: A Radiation Exchange Factor Generator for Rod Bundles, Pacific Northwest Laboratory Technical Report, PNL-6342 (1987). 
  15. W.G. Luscher and K.J. Geelhood. Material Property Correlations: Comparisons Between FRAPCON-3.4, FRAPTRAN-1.4, and MATPRO, Pacific Northwest National Laboratory Technical Report, NUREG/CR-7024, PNNL-19417 (2010). 
  16. EnergySolutions Spent Fuel Division, Inc. FuelSolutionsTM W21 Canister Storage Final Safety Analysis Report, Revision 5, NRC Docket No. 72-1026, EnergySolutions Report, WSNF-221 (2015). 
  17. E.M. Sparrow and A.L. Loeffler Jr., "Longitudinal Laminar Flow Between Cylinders Arranged in Regular Array", AIChE J., 5(3), 325-330 (1959).  https://doi.org/10.1002/aic.690050315
  18. H.C. Hottel and A.F. Saroffim, Radiative Heat Transfer, McGraw-Hill, Inc., New York (1967). 
  19. R. Siegel and J.R. Howell, Thermal Radiation Heat Transfer, 3rd ed., Hemisphere Publishing Corporation, USA (1992). 
  20. O.E. Turgut, M. Asker, and M.T. Coban, "A Review of Non Iterative Friction Factor Correlations for the Calculation of Pressure Drop in Pipe", Bitlis Eren Univ. J. Sci. Technol., 4(1), 1-8 (2014). https://doi.org/10.17678/beujst.90203