Abstract
Zirconium alloys are commonly used as fuel-cladding tubes in water reactors because of their inherent resistance to a variety of environmental conditions. One of the major fuel-reliability issues of the 1970s and early 1980s was pellet cladding interaction (PCI). The mechanism of PCI is one of stress corrosion cracking (SCC) by a combination of aggressive fission products and cladding stress from pellet expansion. The severity of the problem, in particular in boiling water reactors, led to the development of barrier cladding by co-extrusion of Zircaloy-2 with an inner iodide zirconium that essentially eliminated the PCI-related failures. However, the substantially lower corrosion resistance of the zirconium layer led to clad breach and failures by other mechanisms. The difference in corrosion resistance could lead to some dramatic differences in post-failure fuel operations. This article briefly summarizes how PCI-SCC factors led to the development of PCI-resistant fuel cladding and concludes with a note on future research needs.
Article PDF
Similar content being viewed by others
Avoid common mistakes on your manuscript.
References
G.P. Sabol et al., “Development of a Cladding Alloy for High Burnup,” Zirconium in the Nuclear Industry: Eighth Int. Symp., ASTM STP 1023 (Philadelphia, PA: ASTM, 1989), pp. 227–244.
J.P. Mardon, D. Charquet, and J. Senevat, “Development of New Zirconium Alloys for PWR Fuel Rod Cladding,” Int. Topical Meeting on Light Water Reactor Fuel Performance (La Grange Park, IL: ANS, 1994), pp. 643–649.
R.J. Comstock, G. Schoenberger, and G.P. Sabol, “Influence of Processing Variables and Alloy Chemistry on the Corrosion Behavior of ZIRLO@ Nuclear Fuel Cladding,” Zirconium in the Nuclear Industry: Eleventh Int. Symp., ASTM STP 1295 (Philadelphia, PA: ASTM, 1996), pp. 710–725.
K.L. Murty, J. Ravi, and S.T. Mahmood, “Effects of Recrystallization and Nb-Additions on Texture and Mechanical Anisotropy of Zircaloy,” Nuclear Engineering and Design, 148 (1994), pp. 1–15.
K.L. Murty, J. Ravi, and Wiratmo, “Transitions in Creep Mechanisms and Creep Anisotropy in a Zr-1Nb-1Sn-0.2Fe Sheet,” Nuclear Engineering and Design, 156 (1995), pp. 359–371.
K.L. Murty, “Stress Corrosion Cracking and Pellet Cladding Mechanical Interaction of Zircaloys—Application to LWRs,” Emerging Trends in Corrosion Control—Evaluation, Monitoring and Solutions, ed. A.S. Khanna, K.S. Sharma, and A.K. Sinha (New Delhi, India: Akademic Books International, 1999), pp. 702–710.
F. Garzarolli, R. von Jan, and H. Stehle, “The Main Causes of Fuel Element Failure in Water-Cooled Power Reactors,” Atomic Energy Rev., 17 (1) (1979), pp. 31–128.
J.S. Armijo, L.F. Coffin, and H.S. Rosenbaum, “Development of Zirconium-Barrier Fuel Cladding,” Zirconium in the Nuclear Industry: Tenth Int. Symp., ASTM STP 1245 (Philadelphia, PA: ASTM, 1994), pp. 3–18.
H.S. Rosenbaum, “The Interaction of lodine with Zircaloy 2,” Electrochemical Technol., 4 (1966), pp. 153–156.
M. Peehs, H. Stehle, and E. Steinberg, “Out-of-Pile Testing of Iodine Stress Corrosion Cracking in Zircaloy Tubing in Relation to the Pellet-Cladding Interaction Phenomenon,” Zirconium in the Nuclear Industry (Fourth Conference), ASTM STP 681 (Philadelphia, PA: ASTM, 1979), pp. 244–260.
D. Cubicciotti, R.L. Jones, and B.C. Syrett, “Stress Corrosion Cracking of Zircaloys,” EPRI NP-1329 (March 1980).
I.T.A. Roberts, Structural Materials in Nuclear Power Systems (New York: Plenum Press, 1981), p. 66.
D.S. Tomalin, R.B. Adamson, and R.P. Gangloff, “Performance of Irradiated Copper and Zirconium Barrier-Modified Zircaloy Cladding Under Simulated Pellet-Cladding Interaction Conditions,” Zirconium in the Nuclear Industry (Fourth Conference), ASTM STP 681 (Philadelphia, PA: ASTM, 1979), pp. 122–144.
K.L. Murty, “Texture-Based Physical, Mechanical and Corrosion Characteristics of Zirconium Alloys,” Textures in Materials Research, ed. R.K. Ray and A.K. Singh (New Delhi: Oxford & IBH Publishing Co., 1999), pp. 113–160.
I. Shuster and C. Lemaignan, “Influence of Iodine-Induced Stress Corrosion Cracking of Zircaloy-4 Cladding Tubes,” J. Nucl. Mater., 189 (1992), pp. 157–166.
B.L. Adams, D.L. Baty, and K.L. Murty, “A Textural Model for SCC Susceptibility in hcp Metals (Zircaloy),” Scripta Met., 12 (1978), pp. 1151–1155.
K.L. Murty, D.L. Baty, and B.L. Adams, “Texture-Based Computer Modelling of Environmentally Induced Cleavage in HCP Metals,” Proc. of the 10th Int. Conf. on Metallic Corrosion (ICMC-10) (Madras, India: Trans Tech Publications, 1987), pp. 1799–1823.
H.S. Rosenbaum et al., “Zirconium-Barrier Cladding Attributes,” Zirconium in the Nuclear Industry: Seventh Int. Symp., ASTM STP 939 (Philadelphia, PA: ASTM, 1987), pp. 675–699.
A. Jonsson et al., “Failure of a Barrier Rod in Oskarshamn 3,” Proc., Int. Topical Meeting on LWR Fuel Performance (La Grange Park, IL: ANS, 1991), pp. 371–377.
J.H. Davies and G.A. Potts, “Post-Defect Behavior of Barrier Fuel,” Proc., Int. Topical Meeting on LWR Fuel Performance (La Grange Park, IL: ANS, 1991), pp. 272–284.
K.G. Turnage et al., “Fission Product Analysis and Operational Experience with Leaking Fuel Rods at the Hatch Nuclear Unit”, Proc., Int. Topical Meeting on LWR Fuel Performance (La Grange Park, IL: ANS, 1994), pp. 467–476.
K. Edsinger, J.H. Davies, and R.B. Adamson, “Degraded Fuel Cladding Fractography and Fracture Behavior,” Zirconium in the Nuclear Industry: 12th Int. Symp., ASTM STP 1354 (Philadelphia, PA: ASTM, 2000), pp. 316–339.
K. Edsinger, “A Review of Fuel Degradation in BWRs,” Proc. Int. Topical Meeting on LWR Fuel Performance (La Grange Park, IL: ANS, 2000), pp. 523–540.
D. Schrire et al., “Secondary Defect Behavior in ABB BWR Fuel,” Proc. Int. Topical Meeting on LWR Fuel Performance (La Grange Park, IL: ANS, 1994), pp. 398–409.
D.R. Lutz et al., “Effects of Fe on Properties of Zr Barriers,” Proc. of SFEN/ENS Conf. TOPFUEL 99 (Paris: SFEN, 1999), pp. 358–370.
A. Seibold et al., “Fe-Enhanced Zr Liner Cladding,” Proc. Int. Topical Meeting on LWR Fuel Performance (La Grange Park, IL: ANS, 1997), pp. 337–341.
J.S. Armijo, H.S. Rosenbaum, and C.D. Williams, Method for Making Fuel Cladding Having Zirconium Barrier Layers and Inner Liners, U.S. patent 5,383,228 (17 January 1995).
C.D. Williams et al., “Zircaloy-2 Lined Zirconium Barrier Fuel Cladding,” Zirconium in the Nuclear Industry: Eleventh Int. Symp., ASTM STP 1295 (Philadelphia, PA: ASTM, 1996), pp. 676–694.
J.H. Davies, S. Vaidyanathan, and R.A Rand, “Modified UO2 Fuel for High Burnups,” Proc. Of SFEN/ENS Conf. TOPFUEL 99 (Paris: SFEN: 1999), pp 385–395.
Author information
Authors and Affiliations
Additional information
For more information, contact K.L. Murty, North Carolina State University, Department of Nuclear Engineering, Raleigh, NC 27695-7909 USA; (919) 515-3657; fax (919) 515-5115; e-mail murty@eos.ncsu.edu.
Rights and permissions
About this article
Cite this article
Edsinger, K., Murty, K.L. LWR pellet-cladding interactions: Materials solutions to SCC. JOM 53, 9–13 (2001). https://doi.org/10.1007/s11837-001-0079-7
Issue Date:
DOI: https://doi.org/10.1007/s11837-001-0079-7