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Hydride-Related Degradation of Spent-Fuel Cladding Under Repository Conditions

Published online by Cambridge University Press:  10 February 2011

H. M. Chung
H. M. Chung*
Affiliation:
Energy Technology Division, Argonne National Laboratory, Argonne, IL 60439heechung@anl.gov
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Abstract

This report summarizes results of an analysis of hydride-related degradation of commercial spent-nuclear-fuel cladding under repository conditions. Based on applicable laboratory data on critical stress intensity obtained under isothermal conditions, occurrence of delayed hydride cracking from the inner-diameter side of cladding is concluded to be extremely unlikely. The key process for potential initiation of delayed hydride cracking at the outer-diameter side is long-term microstructural evolution near the localized regions of concentrated hydrides, i.e., nucleation, growth, and cracking of hydride blisters. Such locally concentrated hydrides are, however, limited to some high-burnup cladding only, and the potential for crack initiation at the outer-diameter side is expected to be insignificant for the majority of spent fuels. Some degree of hydride reorientation could occur in highburnup spent-fuel cladding. However, even if hydride reorientation occurs, accompanying stress-rupture failure in spent fuel cladding is unlikely to occur.

Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 608: Symposium QQ – Scientific Basis for Nuclear Waste Management XXIII , 1999 , 17
Copyright
Copyright © Materials Research Society 2000

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References

REFERENCES

1 Code of Federal Regulations 1977, Title 10 Part 60, Section 60.113, Article (a)(ii)(B), U.S. Government Printing Office, Washington DC, pp. 137.Google Scholar
2 Coleman, C. E. 1982, in Zr in the Nuclear Industry: 5th Intnl. Symp., ASTM STP 754, Franklin, D. G., ed., ASTM, Philadelphia, p. 393.Google Scholar
3 Simpson, L. A.; and Puls, M. P. 1979, Met. Trans. A, 10A, 1093.Google Scholar
4 Puls, M. P.; Simpson, L. A.; and Dutton, R. 1982, in Fracture Problems and Solutions in the Energy Industry, Pergamon Press, New York, pp. 13–25.Google Scholar
5 Shi, S.-Q.; and Puls, M. P. 1996, in Hydrogen Effects in Materials, Thompson, A. W. and Moody, N. R., eds., TMS, Warrendale, PA, p. 612.Google Scholar
6 Efsing, P.; and Petterson, K. 1996, in Zr in the Nuclear Industry: 11th Intnl. Symp., ASTM STP 1295, Bradley, E. R. and Sabol, G. P., eds., ASTM, Philadelphia, pp. 394.Google Scholar
7 Chow, C. K.; and Simpson, L. A. 1986, in Case Histories Involving Fatigue and Fracture Mechanics, ASTM STP 918, ASTM, p. 78.Google Scholar
8 Cheadle, B. A.; Coleman, C. E.; and Ambler, J. F. R. 1987, in Zr in the Nuclear Industry: 7th Intnl. Symp., ASTM STP 939, Adamson, R. B. and Swam, L. F. P. Van, eds., ASTM, Philadelphia, pp. 224–240.Google Scholar
9 Leger, M.; Moan, G. D.; Wallace, A. C.; and Watson, N. J. 1989, in Zr in the Nuclear Industry: 8th Intnl. Symp., ASTM STP 1023, Swam, L. F. P. van and Eucken, C. M., eds., ASTM, Philadelphia, pp. 50–65.Google Scholar
10 Coleman, C. E.; Cheadle, B. A.; Causey, A. R.; Chow, C. K.; Davies, P. H.; McManus, M. D.; Rodgers, D.; Sagat, S.; van Drunen, G. 1989, ibid., pp. 35–49.Google Scholar
11 Moan, G. D.; Coleman, C. E.; Price, E. G.; Rodgers, D. K.; and Sagat, S. 1990, Int. J. Pres. Vessel & Piping, 43, 1–21.Google Scholar
12 Einziger, R. E.; and Kohli, R. 1984, Nucl. Technol. 67, 107–123.Google Scholar
13 Smith, G. P. Jr; Pirek, R. C.; Freeburn, H. R.; and Schrire, D. Jr 1994, The Evaluation and Demonstration of Methods for Improved Nuclear Fuel Utilization, DOE/ET/34013-15, CEND-432, ABB Combustion Engineering, pp. 4-60 to 4-73.Google Scholar
14 Garde, A. M.; Smith, G. P.; and Pirek, R. C. 1996, in Zr in the Nuclear Industry: 11 th Intnl. Symp., ASTM STP 1295, Bradley, E. R. and Sabol, G. P., eds., ASTM, Philadelphia, pp. 407–430.Google Scholar
15 Yang, R. L.; Ozer, O.; and Klepfer, H. H. 1991, in Proc. Intnl. Topical Meeting on LWR Fuel Performance, April 21-24, 1991, Avignon, France, ANS and ENS, pp. 258–271.Google Scholar
16 Guedeney, P.; Trotabas, M.; Boschiero, M.; Forat, C.; and Blanpain, P. 1991, in Proc. Intnl. Topical Meeting on LWR Fuel Performance, April 21-24, 1991, Avignon, France, ANS and ENS, pp. 627–638.Google Scholar
17 McMinn, A.; Darby, E. C.; and Schofield, J. S. 1998, in Zr in the Nuclear Industry: 12th Intnl. Symp., June 15-18, 1998, Toronto, in press.Google Scholar
18 Marshall, R. P. 1967, J. Nucl. Mater. 24, 34–48.Google Scholar
19 Bai, J. B.; Ji, N.; Gilbon, D.; Prioul, C.; and Francois, D. 1994, Met. and Mater.Trans. A, 25A, 1199–1208.Google Scholar
20 Chan, K. S. 1996, J. Nucl. Mater. 227, 220–236.Google Scholar
21 Hardie, D.; and Shanahan, M. W. 1975, J. Nucl. Mater. 55, 1–13.Google Scholar
22 Chung, H. M.; Yaggee, F. L.; and Kassner, T. F. 1987, in Zr in the Nuclear Industry: 7th Intnl. Symp., ASTM STP 939, Adamson, R. B. and Swam, L. F. P. Van, eds., ASTM, Philadelphia, pp. 775–801.Google Scholar
23 Garde, A. M. 1989, in Zr in the Nuclear Industry: 8th Intnl. Symp., ASTM STP 1023, Swam, L. F. P. Van and Eucken, C. M., eds., ASTM, Philadelphia, pp. 548–569.Google Scholar