Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Methodology for numerical evaluation of fracture resistance under pinch loading of spent nuclear fuel cladding containing reoriented hydrides

  • Seyeon Kim (Department of Mechanical Engineering, Keimyung University) ;
  • Sanghoon Lee (Department of Mechanical Engineering, Keimyung University)
  • Received : 2023.11.22
  • Accepted : 2024.01.02
  • Published : 2024.06.25
Download PDF
⟨ Previous Next ⟩

Abstract

It is important to maintain cladding integrity in spent nuclear fuel management. This study proposes a numerical analysis method to evaluate the fracture resistance of irradiated zirconium alloy cladding under pinch load known to cause Mode-III failure. The mechanical behavior and fracture of the cladding under pinch loading can be evaluated by a Ring Compression Test (RCT). To simulate the fracture of hydride precipitates, zirconium matrix, and Zr/hydride interfaces under the stress field generated by RCT, a micro-structure crack propagation simulation method based on Continuum Damage Mechanics (CDM) has been proposed. Our RCT simulation model was constructed from microscopic images of irradiated cladding. In this study, we developed an automated process to generate a pixel-based finite element model by separating the hydride precipitates, zirconium matrix, and interfaces using an image segmentation method. The appropriate element size was selected to ensure the efficiency and accuracy of a crack propagation simulation. The load-displacement curves and strain energies from RCT were compared and analyzed with the simulation results of different element sizes. The finalized RCT simulation model can be used to establish the failure criterion of fuel rods under pinch loading. The advantages and limitations of the proposed method are fully discussed here.

Keywords

Acknowledgement

This work was funded by the Nuclear Safety Research Program through the Korea Foundation of Nuclear Safety (KoFONS) using the financial resource granted by the Nuclear Safety and Security Commission (NSSC) of the Republic of Korea (No. 2106042).

References

  1. U. S. Nuclear Regulatory Commission, Spent Nuclear Fuel Transportation Risk Assessment-Final Report (NUREG-2125), US Nuclear Regulatory Commission, Washington, DC, 2014.
  2. U. S. Nuclear Regulatory Commission, "NUREG-1864, 'A pilot probabilistic risk assessment of a dry storage system at a nuclear power plant.'" [Online]. Available: www.nrc.aov/readina-rmldoc-collectionslnureas.
  3. U. S. Nuclear Regulatory Commission, Spent fuel transportation risk assessment final report office of nuclear materials safety and safeguards [Online]. Available: http://www.nrc.gov/reading-rm.html.
  4. T.L. Sanders, et al., A Method for Determining the Spent-Fuel Contribution to Transport Cask Containment Requirements, 1992.
  5. Office of nuclear material safety and safeguards dry storage and transportation of high burnup spent nuclear fuel final report [Online]. Available, www.nrc.gov/reading-rm.html.
  6. J.J. Kearns, C.R. Woods, Effect of texture, grain size, and cold work on the precipitation of oriented hydrides in Zircaloy tubing and plate, J. Nucl. Mater. 20 (3) (1966) 241-261, https://doi.org/10.1016/0022-3115(66)90036-5.
  7. Y.-J. Kim, D.-H. Kook, T.-H. Kim, J.-S. Kim, Stress and temperature-dependent hydride reorientation of Zircaloy-4 cladding and its effect on the ductility degradation, J. Nucl. Sci. Technol. 52 (5) (2015) 717-727, https://doi.org/10.1080/00223131.2014.978829.
  8. A.M. Alam, C. Hellwig, Cladding tube deformation test for stress reorientation of hydrides, J. ASTM Int. (JAI) 5 (2) (2008), https://doi.org/10.1520/JAI101110.
  9. M.C. Billone, T.A. Burtseva, R.E. Einziger, Ductile-to-brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions, J. Nucl. Mater. 433 (1-3) (2013) 431-448, https://doi.org/10.1016/j.jnucmat.2012.10.002.
  10. Ductile-Brittle Transition Temperature for High Burnup Zircaloy-4 and ZIRLOTM Cladding Alloys Exposed to Simulated Drying Storage Conditions", Nuclear Engineering Division. [Online]. Available: www.anl.gov.
  11. M.C. Billone, T.A. Burtseva, M.A. Martin-Rengel, Effects of Lower Drying-Storage Temperatures on the DBTT of High-Burnup PWR Cladding, Argonne National Lab. (ANL), Argonne, IL (United States), 2015.
  12. C.E. Ells, Hydride Precipitates in Zirconium Alloys (a review), J. Nucl. Mater. 28 (2) (1968) 129-151.
  13. A.T. Motta, et al., Hydrogen in zirconium alloys: a review, J. Nucl. Mater. 518 (2019) 440-460, https://doi.org/10.1016/j.jnucmat.2019.02.042.
  14. S. Suman, M.K. Khan, M. Pathak, R.N. Singh, J.K. Chakravartty, Hydrogen in zircaloy: mechanism and its impacts, Int. J. Hydrogen Energy 40 (17) (2015) 5976-5994, https://doi.org/10.1016/j.ijhydene.2015.03.049. Elsevier Ltd.
  15. G.J.C. Carpenter, The Dilatational Misfit of Zirconium Hydrides Precipitated in Zirconium, J. Nucl. Mater. 48 (3) (1973) 264-266.
  16. V. Perovic, G.C. Weatherly, C.J. Simpson, Hydride precipitation in α/β zirconium alloys, Acta Metall. 31 (9) (Sep. 1983) 1381-1391, https://doi.org/10.1016/0001-6160(83)90008-1.
  17. S. Suman, M.K. Khan, M. Pathak, R.N. Singh, Investigation of elevated-temperature mechanical properties of δ-hydride precipitate in Zircaloy-4 fuel cladding tubes using nanoindentation, J. Alloys Compd. 726 (2017) 107-113, https://doi.org/10.1016/j.jallcom.2017.07.321.
  18. A. Needleman, A Continuum Model for Void Nucleation by Inclusion Debonding, 1987.
  19. D.C. Ahn, P. Sofronis, R. Dodds, Modeling of hydrogen-assisted ductile crack propagation in metals and alloys, Int. J. Fract. 145 (2007) 135-157.
  20. H. Chan, S.G. Roberts, J. Gong, Micro-scale fracture experiments on zirconium hydrides and phase boundaries, J. Nucl. Mater. 475 (2016) 105-112.
  21. M. Le Saux, J. Besson, S. Carassou, C. Poussard, X. Averty, Behavior and failure of uniformly hydrided Zircaloy-4 fuel claddings between 25 ℃ and 480 ℃ under various stress states, including RIA loading conditions, Eng. Fail. Anal. 17 (3) (Apr. 2010) 683-700, https://doi.org/10.1016/j.engfailanal.2009.07.001.
  22. J. Lemaitre, J.-L. Chaboche, Aspect phenomenologique de la rupture par endommagement, J. Mec. Appl. 2 (3) (1978).
  23. S. Murakami, Notion of Continuum Damage Mechanics and its Application to Anisotropic Creep Damage Theory, 1983.
  24. K.S. Chan, A fracture model for hydride-induced embrittlement, Acta Metall. Mater. 43 (12) (1995) 4325-4335.
  25. K.-F. Nilsson, N. Jaksic, V. Vokal, An elasto-plastic fracture mechanics based model for assessment of hydride embrittlement in zircaloy cladding tubes, J. Nucl. Mater. 396 (1) (2010) 71-85.
  26. K.S. Chan, A micromechanical model for predicting hydride embrittlement in nuclear fuel cladding material, J. Nucl. Mater. 227 (3) (1996) 220-236.
  27. P. Germain, Cours de mecanique des milieux continus, no. V. 1, Masson, 1973 [Online]. Available: https://books.google.co.kr/books?id=Y8i2AQAACAAJ.
  28. J. Janson, J. Hult, Damage mechanics and fracture mechanics: a combined approach, J. Mec. Appl. 1 (1977) 69-84.
  29. M. Smith, ABAQUS/standard User's Manual, version 6.9, 2009.
  30. A. Hillerborg, M. Modeer, P.-E. Petersson, Analysis of crack formation and crack growth in concrete by means of fracture mechanics and finite elements, Cement Concr. Res. 6 (6) (1976) 773-781.
  31. J. Serra, Vol. 2 (1988), Image Analysis and Mathematical Morphology, vol. 1, 1982 [Online]. Available: citeulike-article-id:7544267. (Accessed 5 July 2023).
  32. E. Fogel, D. Halperin, Exact and efficient construction of Minkowski sums of convex polyhedra with applications, Comput. Aided Des. 39 (11) (2007) 929-940.
  33. R. Gardner, The brunn-minkowski inequality, Bull. Am. Math. Soc. 39 (3) (2002) 355-405.
  34. R. Van Den Boomgaard, R. Van Balen, Methods for fast morphological image transforms using bitmapped binary images, CVGIP Graph. Models Image Process. 54 (3) (1992) 252-258.
  35. E.R. Dougherty, An introduction to morphological image processing, in: SPIE, Optical Engineering Press, 1992.
  36. P.K. Ghosh, A unified computational framework for Minkowski operations, Comput. Graph. 17 (4) (1993) 357-378.
  37. K.J. Geelhood, C.E. Beyer, W.G. Luscher, PNNL Stress/strain Correlation for Zircaloy, Pacific Northwest National Lab.(PNNL), Richland, WA (United States), 2008.
  38. M.P. Puls, S.-Q. Shi, J. Rabier, Experimental studies of mechanical properties of solid zirconium hydrides, J. Nucl. Mater. 336 (1) (2005) 73-80.
  39. A. Rico, M.A. Martin-Rengel, J. Ruiz-Hervias, J. Rodriguez, F.J. Gomez-Sanchez, Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding, J. Nucl. Mater. 452 (1-3) (2014) 69-76.
  40. S. Yamanaka, et al., Thermal and mechanical properties of zirconium hydride, J. Alloys Compd. 293 (1999) 23-29.
  41. M. Kuroda, K. Yoshioka, S. Yamanaka, H. Anada, F. Nagase, H. Uetsuka, Influence of precipitated hydride on the fracture behavior of Zircaloy fuel cladding tube, J. Nucl. Sci. Technol. 37 (8) (2000) 670-675.
  42. C. Evans, Micromechanisms and micromechanics of Zircaloy-4 (2014).
  43. K. Kese, P.A.T. Olsson, A.-M.A. Holston, E. Broitman, High temperature nanoindentation hardness and Young's modulus measurements in a neutron-irradiated fuel cladding material, J. Nucl. Mater. 487 (2017) 113-120.
  44. W. Zhu, R. Wang, G. Shu, P. Wu, H. Xiao, First-principles study of different polymorphs of crystalline zirconium hydride, J. Phys. Chem. C 114 (50) (2010) 22361-22368.
  45. P.A.T. Olsson, A.R. Massih, J. Blomqvist, A.-M.A. Holston, C. Bjerk'en, Ab initio thermodynamics of zirconium hydrides and deuterides, Comput. Mater. Sci. 86 (2014) 211-222.
  46. P.F. Weck, E. Kim, V. Tikare, J.A. Mitchell, Mechanical properties of zirconium alloys and zirconium hydrides predicted from density functional perturbation theory, Dalton Trans. 44 (43) (2015) 18769-18779.
  47. S. Suman, M.K. Khan, M. Pathak, R.N. Singh, Investigation of elevated-temperature mechanical properties of δ-hydride precipitate in Zircaloy-4 fuel cladding tubes using nanoindentation, J. Alloys Compd. 726 (2017) 107-113.