- Volume 41 Issue 2
DOI QR Code
DELAYED HYDRIDE CRACKING IN ZIRCALOY FUEL CLADDING - AN IAEA COORDINATED RESEARCH PROGRAMME
- Coleman, C. (AECL, Chalk River Laboratories) ;
- Grigoriev, V. (Studsvik Nuclear AB) ;
- Inozemtsev, V. (IAEA, Nuclear Fuel Cycle and Materials Section, Division of Nuclear Fuel Cycle and Waste Technology, Department of Nuclear Energy) ;
- Markelov, V. (VNIINM, A. A. Bochvar All Russia Research Institute of Inorganic Materials) ;
- Roth, M. (Institute for Nuclear Research) ;
- Makarevicius, V. (Lithuanian Energy Institute) ;
- Kim, Y.S. (Korea Atomic Energy Research Institute) ;
- Ali, Kanwar Liagat (Pakistan Institute of Nuclear Science and Technology) ;
- Chakravartty, J.K. (Department of Atomic Energy, Bhabha Atomic Research Centre, Materials Science Division) ;
- Mizrahi, R. (Comision Nacional de Energia Atomica) ;
- Lalgudi, R. (Instituto de Pesquisas Energeticas e Nucleares)
- Published : 2009.03.30
The rate of delayed hydride cracking (DHC), V, has been measured in cold-worked and stress-relieved Zircaloy-4 fuel cladding using the Pin-Loading Tension technique. At
- E. C. W. Perryman, "Pickering Pressure Tube Cracking Experience", Nuclear Energy, 17, 95-105, (1978).
- D. Schrire, B. Grapengiesser, L. Hallstadius, L. Lundholm, G. Lysell, G. Frenning, G. Ronnberg, and A. Jonsson, "Secondary Defect Behaviour in ABB BWR Fuel", Proc. International Topical Meeting on Light Water Reactor Fuel Performance, ANS, West Palm Beach, 398-409, (1994).
- K. Edsinger, "A Review of Fuel Degradation in BWRs", Proc. International Topical Meeting on Light Water Reactor Fuel Performance, ANS, Park City, USA, 162-179, (2000).
- C. E. Coleman and V. V. Inozemstev, "Measurement of Rates of Delayed Hydride Cracking (DHC) in Zr-2.5Nb - An IAEA Coordinated Research Project", J. ASTM International, 5 (2), Paper ID: 101091, (2008). https://doi.org/10.1520/JAI101091
- P. Efsing and K. Pettersson, "The Influence of Temperature and Yield Strength on Delayed Hydride Cracking in Hydrided Zircaloy-2", Zirconium in the Nuclear Industry: Eleventh International Symposium", ASTM STP 1295, 394-404, (1996). https://doi.org/10.1520/STP16182S
- K. Edsinger, J. H. Davies and R. B. Adamson, "Degraded Fuel Cladding Fractography and Fracture Behaviour", Zirconium in the Nuclear Industry: Twelfth International Symposium", ASTM STP 1354, 316-339, (2000). https://doi.org/10.1520/STP14306S
- V. Grigoriev and R. Jakobsson, "Delayed Hydrogen Cracking velocity and J-Integral Measurements on Irradiated BWR Cladding", Zirconium in the Nuclear Industry: Fourteenth International Symposium", ASTM STP 1467, 711-728, (2006). https://doi.org/10.1520/JAI12434
- F. H. Huang and W. J. Mills, "Delayed Hydride Cracking Behavior for Zircaloy-2 Tubing", Metallurgical Transactions, 22A, 2049-2060, (1991). https://doi.org/10.1007/BF02669872
- M. P. Puls, L. A. Simpson and R. Dutton, "Hydride-induced Crack Growth in Zirconium Alloys", AECL Report, AECL-7392, (1982).
- J. Y. Oh, I. S. Kim and Y. S. Kim, "A Normalization Method for Relationship between Yield Stress and Delayed Hydride Cracking Velocity in Zr-2.5Nb Alloy", J. Nuclear Science and Technology, 37, 595-600, (2000). https://doi.org/10.3327/jnst.37.595
- A. Sawatzky, "The Diffusivity and Solubility of Hydrogen in the Alpha-phase of Zircaloy", J. Nuclear Materials, 2, 62-68, (1960). https://doi.org/10.1016/0022-3115(60)90025-8
- A. Sawatzky, G. A. Ledoux, R. L. Tough, C. D. Cann, "Hydrogen Diffusion in Zirconium-Niobium Alloys, Proc. International Symposium on Metal-Hydrogen Systems, Pergamon Press, 109-120, (1981).
- J. J. Kearns, "Terminal solubility and Partitioning of hydrogen in the Alpha Phase of Zirconium, Zircaloy-2 and Zircaloy-4, J. Nuclear Materials, 22, 292-303, (1967). https://doi.org/10.1016/0022-3115(67)90047-5
- S. T. Mahmood, D .M. Farkas, R. B. Adamson and Y. Etoh, "Post-Irradiation Characterization of Ultra-High-Fluence Zircaloy-2 Plate," Zirconium in the Nuclear Industry: Twelfth International Symposium, ASTM STP 1354, 139-169, (2000). https://doi.org/10.1520/STP14299S
- R. R. Smith and R. L.Eadie, "High Temperature Limit for Delayed Hydride Cracking", Scripta Metallurgica, 22, 833-836, (1988). https://doi.org/10.1016/S0036-9748(88)80058-9
- S. Sagat and M. P. Puls, "Temperature Limit for Delayed Hydride Cracking in Zr-2.Nb Alloys," 17th Inter. Conf. Structural Mechanics in Reactor Technology, Paper G06-4, (2003).
- M. Resta Levi and M. P. Puls, "DHC Behaviour of Irradiated Zr-2.5Nb Pressure Tubes up to 365°C," 18th Inter. Conf. Structural Mechanics in Reactor Technology, Paper G10-3, (2005).
- J. F. R. Ambler, "Effect of Direction of Approach to Temperature on the Delayed Hydride Cracking Behaviour of Cold-worked Zr-2.5Nb," Zirconium in the Nuclear Industry: Sixth International Symposium, ASTM STP 824, 653-674, (1984). https://doi.org/10.1520/STP34499S
- S. Sagat, C. E. Coleman, M. Griffiths and B. J. S. Wilkins, "Effect of Fluence and Irradiation Temperature on Delayed Hydride Cracking in Zr-2.5Nb," Zirconium in the Nuclear Industry: Tenth International Symposium, ASTM STP 1245, 35-61, (1994). https://doi.org/10.1520/STP15183S
- S. Sagat, C. K. Chow, M. P. Puls and C. E. Coleman, "Delayed Hydride Cracking in a Temperature Gradient," J. Nuclear Materials, 279, 107-117, (2000). https://doi.org/10.1016/S0022-3115(99)00265-2
- Delayed hydride cracking of zirconium alloys: Appearance conditions and basic laws vol.2011, pp.4, 2011, https://doi.org/10.1134/S0036029511040124
- In Situ Scanning Electron Microscope Observation and Finite Element Method Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes vol.8, pp.3, 2011, https://doi.org/10.1520/JAI102960
- Temperature dependences of the delayed hydride cracking rate of fuel claddings made of zirconium alloys of various compositions vol.2014, pp.4, 2014, https://doi.org/10.1134/S0036029514040077
- Phase structural ordering kinetics of second-phase formation in the vicinity of a crack pp.1573-2673, 2017, https://doi.org/10.1007/s10704-017-0242-y
- The Effect of Microstructure on Delayed Hydride Cracking Behavior of Zircaloy-4 Fuel Cladding—An International Atomic Energy Agency Coordinated Research Program vol.7, pp.5, 2010, https://doi.org/10.1520/JAI103008