Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Development and testing of the hydrogen behavior tool for Falcon - HYPE

  • Piotr Konarski (Laboratory for Reactor Physics and Thermal-Hydraulics, Paul Scherrer Institut (PSI)) ;
  • Cedric Cozzo (Laboratory for Reactor Physics and Thermal-Hydraulics, Paul Scherrer Institut (PSI)) ;
  • Grigori Khvostov (Laboratory for Reactor Physics and Thermal-Hydraulics, Paul Scherrer Institut (PSI)) ;
  • Hakim Ferroukhi (Laboratory for Reactor Physics and Thermal-Hydraulics, Paul Scherrer Institut (PSI))
  • Received : 2023.07.03
  • Accepted : 2023.11.07
  • Published : 2024.02.25
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Abstract

The presence of hydrogen absorbed by zirconium-based cladding materials during reactor operation can trigger degradation mechanisms and endanger the rod integrity. Ensuring the durability of the rods in extended time-frames like dry storage requires anticipating hydrogen behavior using numerical modeling. In this context, the present paper describes a hydrogen post-processing tool for Falcon - HYPE, a PSI's in-house tool able to calculate hydrogen uptake, transport, thermochemistry, reorientation of hydrides and hydrogen-related failure criteria. The tool extracts all necessary data from a Falcon output file; therefore, it can be considered loosely coupled to Falcon. HYPE has been successfully validated against experimental data and applied to reactor operation and interim storage scenarios to present its capabilities.

Keywords

Acknowledgement

This work was partly funded by the Swiss Nuclear Safety Inspectorate ENSI as part of the DRYstars project (CTR00521) conducted within the framework of the STARS program (http://www.psi.ch/stars). The authors would also like to acknowledge the PSI Laboratory for Nuclear Materials for fruitful discussions.

References

  1. S. Yagnik, Fuel analysis and licensing code: falcon MOD01: verification and validation, EPRI technical report 3002005391 (2015). 
  2. P. Konarski, C. Cozzo, G. Khvostov, H. Ferroukhi, Extension of Falcon's modelling capabilities to dry storage: analysis of cladding creep and helium-induced swelling, The Nuclear Materials Conference NuMat (2020). 
  3. P. Konarski, C. Cozzo, G. Khvostov, H. Ferroukhi, Dry storage modeling activities at PSI: implementation and testing of a creep model for dry storage, Kerntechnik 86 (2020) 419, https://doi.org/10.3139/124.200072. 
  4. P. Konarski, C. Cozzo, G. Khvostov, H. Ferroukhi, Spent nuclear fuel in dry storage conditions - current trends in fuel performance modeling, J. Nucl. Mater. 555 (2021), 153138, https://doi.org/10.1016/j.jnucmat.2021.153138. 
  5. P. Konarski, C. Cozzo, G. Khvostov, H. Ferroukhi, Modeling of hydrogen behavior in liner claddings, J. Nucl. Mater. 573 (2023), 154125, https://doi.org/10.1016/j.jnucmat.2022.154125. 
  6. A. Motta, L. Capolungo, L.-Q. Chen, M. Cinbiz, M. Daymond, D. Koss, E. Lacroix, G. Pastore, P.-C. Simon, M. Tonks, B. Wirth, M. Zikryi, Hydrogen in zirconium alloys, A review Journal of Nuclear Materials 518 (2019) 440-460, https://doi.org/10.1016/j.jnucmat.2019.02.042. 
  7. O. Courty, A. Motta, J. Hales, Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding, J. Nucl. Mater. 452 (2014) 311-320, https://doi.org/10.1016/j.jnucmat.2014.05.013. 
  8. D. Stafford, Multidimensional simulations of hydrides during fuel rod lifecycle, J. Nucl. Mater. 466 (2015) 362-372, https://doi.org/10.1016/j.jnucmat.2015.06.037. 
  9. F. Passelaigue, E. Lacroix, G. Pastore, A. Motta, Implementation and validation of the hydride nucleation-growth-dissolution (HNGD) model in BISON, J. Nucl. Mater. 544 (2021) 152683, https://doi.org/10.1016/j.jnucmat.2020.152683. 
  10. F. Passelaigue, E. Lacroix, G. Pastore, A. Motta, Predicting the hydride rim by improving the solubility limits in the Hydride Nucleation-Growth-Dissolution (HNGD) model, J. Nucl. Mater. 558 (2022) 153363, https://doi.org/10.1016/j.jnucmat.2021.153363. 
  11. F. Feria, L. Herranz, Effect of the oxidation front penetration on in-clad hydrogen migration, J. Nucl. Mater. 500 (2018) 349-360, https://doi.org/10.1016/j.jnucmat.2018.01.011. 
  12. F. Feria, C. Aguado, L. Herranz, Extension of FRAPCON-xt to hydride radial reorientation in dry storage, Ann. Nucl. Energy 145 (2020) 107559, https://doi.org/10.1016/j.anucene.2020.107559. 
  13. F. Feria, L. Herranz, Assessment of hydride precipitation modelling across fuel cladding: hydriding in non-defective and defective fuel rods, Ann. Nucl. Energy 188 (2023) 109810, https://doi.org/10.1016/j.anucene.2023.109810. 
  14. M. Sugisaki, K. Hashizume, Y. Hatano, Estimation of hydrogen redistribution in zircaloy cladding of spent fuel under thermal conditions of dry storage and evaluation of its influence on mechanical properties of the cladding, IAEA-TECDOC-1316 (2002). 
  15. A. Aly, K. Gamble, M. Avramova, R. Williamson, K. Ivanov, Modeling 3D hydrogen diffusion and localized hydride formation in zirconium alloy cladding using high fidelity multi-physics coupled codes, in: M&C 2017 - International Conference on Mathematics & Computational Methods Applied to Nuclear Science & Engineering, Jeju, Korea, 2017. 
  16. C. Lee, Y. Lee, Simulation of hydrogen diffusion along the axial direction in zirconium cladding tube during dry storage, J. Nucl. Mater. 579 (2023) 154352, https://doi.org/10.1016/j.jnucmat.2023.154352. 
  17. US NRC Spent Fuel Project Office, Interim Staff Guidance-11, Revision 2, 2002. https://www.nrc.gov/reading-rm/doc-collections/isg/isg-11R3.pdf. 
  18. EPRI program 41.03.01: Used Fuel and High-Level Waste Management https://www.epri.com/research/programs/061149/hbudemo. 
  19. J.J. Kearns, Diffusion coefficient of hydrogen in alpha zirconium, Zircaloy-2 and Zircaloy-4, J. Nucl. Mater. 43 (1972) 330, https://doi.org/10.1016/0022-3115(72)90065-7. 
  20. S. Kang, P.-H. Huang, V. Petrov, A. Manera, T. Ahn, B. Kammenzind, A. Motta, Determination of the hydrogen heat of transport in Zircaloy-4, J. Nucl. Mater. 573 (2023) 154122, https://doi.org/10.1016/j.jnucmat.2022.154122. 
  21. W.J. Kammenzind, B. Franklin, D.G. Peters, H.R. Duffin, Hydrogen pickup and redistribution in alpha-annealed zircaloy-4, in: 11th International Symposium on Zirconium in the Nuclear Industry, ASTM STP1295, 1997, p. 338, https://doi.org/10.2172/10181237. 
  22. A. Sawatzky, Hydrogen in zircaloy-2: its distribution and heat of transport, J. Nucl. Mater. 2 (1960) 321-328, https://doi.org/10.1016/0022-3115(60)90004-0. 
  23. P. Konarski, Thermo-chemical-mechanical Modelling of Nuclear Fuel Behavior. Impact of Oxygen Transport in the Fuel on Pellet-Cladding Interaction Materials, Universite de Lyon, 2019. English. NNT: 2019LYSEI080. tel-02570386. 
  24. 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 (2013) 431-448, https://doi.org/10.1016/j.jnucmat.2012.10.002. 
  25. W. Gong, P. Trtik, S. Valance, J. Bertsch, Hydrogen diffusion under stress in Zircaloy: high-resolution neutron radiography and finite element modeling, J. Nucl. Mater. 508 (2018) 459, https://doi.org/10.1016/j.jnucmat.2018.05.079. 
  26. C. Cozzo, G. Khvostov, I. Clifford, H. Ferroukhi, Appraisal of the NRC H-uptake calculation for Swiss boiling water reactors, Nucl. Eng. Des. 391 (2022) 111731, https://doi.org/10.1016/j.nucengdes.2022.111731. 
  27. P. M. Clifford, Acceptable Fuel Cladding Hydrogen Uptake Models US NRC Memorandum, ML15133A306. 
  28. F. Boldt, Implementation of hydrogen solid solubility data and precipitation threshold stresses in the fuel rod code TESPA-ROD, J. Nucl. Eng. Radiat. Sci. 5 (2019) 20904, https://doi.org/10.1115/1.4042118. 
  29. O. Zanellato, M. Preuss, J.-Y. Buffiere, F. Ribeiro, A. Steuwer, J. Desquines, J. Andrieux, B. Krebs, Synchrotron diffraction study of dissolution and precipitation kinetics of hydrides in Zircaloy-4, J. Nucl. Mater. 420 (2012) 537-547, https://doi.org/10.1016/j.jnucmat.2011.11.009. 
  30. Z. Pan, I. Ritchie, M. Puls, The terminal solid solubility of hydrogen and deuterium in Zr-2.5Nb alloys, J. Nucl. Mater. 228 (1996) 227-237, https://doi.org/10.1016/S0022-3115(95)00217-0. 
  31. K. Une, S. Ishimoto, Y. Etoh, K. Ito, K. Ogata, T. Baba, K. Kamimura, Y. Kobayashi, The terminal solid solubility of hydrogen in irradiated Zircaloy-2 and microscopic modeling of hydride behaviour, J. Nucl. Mater. 389 (2009) 127-136, https://doi.org/10.1016/j.jnucmat.2009.01.017. 
  32. A. McMinn, E.C. Darby, J.S. Schofield, The terminal solid solubility of hydrogen in zirconium alloys, Zirconium in the Nuclear Industry: 12th International Symposium STP1354 (2000), https://doi.org/10.1520/STP14300S. 
  33. K. Colas, Fundamental Experiments on Hydride Reorientation in Zircaloy, PhD thesis, The Pennsylvania State University, 2012. 
  34. J. Desquines, D. Drouan, M. Billone, M. Puls, P. March, S. Fourgeaud, C. Getrey, V. Elbaz, M. Philippe, Influence of temperature and hydrogen content on stress-induced radial hydride precipitation in Zircaloy-4 cladding, J. Nucl. Mater. 453 (2014) 131-150, https://doi.org/10.1016/j.jnucmat.2014.06.049. 
  35. Delayed Hydride Cracking Considerations Relevant to Spent Nuclear Fuel Storage, EPRI Technical Report, 2011, 1022921. 
  36. P. Raynaud, R. Einziger, Cladding stress during extended storage of high burnup spent nuclear fuel, J. Nucl. Mater. 464 (2015) 304-312, https://doi.org/10.1016/j.jnucmat.2015.05.008. 
  37. S.Q. Shi, M.P. Puls, Criteria for fracture initiation at hydrides in zirconium alloys-I. Sharp crack tip, J. Nucl. Mater. 208 (1994) 232-242, https://doi.org/10.1016/0022-3115(94)90332-8. 
  38. T. Aliev, M. Kolesnik, Analytical approach to DHC description in zirconium alloys, Int. J. Fract. 228 (2021) 71-84, https://doi.org/10.1007/s10704-021-00515-0. 
  39. S. Sagat, C. Coleman, M. Griffits, B. Wilkins, The effect of fluence and irradiation temperature on delayed hydride cracking in Zr-2.5Nb, in: Zirconium in the Nuclear Industry: Tenth International Symposium, STP15183S, 1994, pp. 35-61, https://doi.org/10.1520/STP15183S. 
  40. Y. Kim, Y. Cheong, Anisotropic delayed hydride cracking velocity of CANDU Zr-2.5Nb pressure tubes, J. Nucl. Mater. 373 (2008) 179-185, https://doi.org/10.1016/j.jnucmat.2007.05.047. 
  41. M. Levi, M. Puls, DHC behaviour of irradiated Zr-2.5Nb pressure tubes up to 365℃, in: Conference on Structural Mechanics in Reactor Technology (SMiRT 18), vol. 18, SMiRT, 2005, pp. 1811-1823. 
  42. F. Huang, W. Mills, Delayed hydride cracking behavior for ZIRCALOY-2 tubing, Metall. Trans. A 22 (1991) 2049-2060, https://doi.org/10.1007/BF02669872. 
  43. Evaluation of Conditions for Hydrogen Induced Degradation of Zirconium Alloys during Fuel Operation and Storage IAEA Technical Report, IAEA-TECDOC-1781. 
  44. J. Hong, M. Park, A. Alvarez Holston, J. Stjarnsater, D. Kook, Threshold stress intensity factor of delayed hydride cracking in irradiated and unirradiated zircaloy-4 cladding, J. Nucl. Mater. 543 (2021) 152596, https://doi.org/10.1016/j.jnucmat.2020.152596. 
  45. S. Wilks, Determination of sample sizes for setting tolerance limits, Ann. Math. Stat. 12 (1941) 91-96, https://doi.org/10.1214/aoms/1177731788. 
  46. J. Markowitz, The thermal diffusion of hydrogen in alpha-delta Zircaloy 2, Trans. Metall. Soc. AIME 221 (1961) 819. 
  47. J. Merlino, Experiments in Hydrogen Distribution in Thermal Gradients Calculated Using BISON M.Eng. Paper in Nuclear Engineering, The Pennsylvania State University, 2019. https://github.com/FloPasselaigue/HNGD/blob/master/Merlino_PSU_MEng_2019.pdf. 
  48. F. Boldt, M. Stuke, M. Peridis, Benchmark on Thermomechanical Fuel Rod Behaviour - Phase I Report GRS Report, GRS-671, 2022. https://www.grs.de/sites/default/files/2022-05/GRS-671.pdf. 
  49. Fuel modelling at extended burnup (FUMEX-II), in: IAEA Coordinated Research Project, IAEA-TECDOC-1687, 2012. https://www-pub.iaea.org/MTCD/Publications/PDF/TE_1687_web.pdf. 
  50. R.L. Eadie, K. Tashiro, D. Harrington, M. Leger, The determination of the partial molar volume of hydrogen in zirconium in a simple stress gradient using comparative microcalorimetry, Scripta Metall. Mater. 26 (1992) 231-236, https://doi.org/10.1016/0956-716X(92)90178-H.