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A Probabilistic-Micromechanical Methodology for Assessing Zirconium Alloy Cladding Failure

Published online by Cambridge University Press:  19 October 2011

Yi-Ming Pan,
K. S. Chan and
D. S. Riha
Yi-Ming Pan
Affiliation:
ypan@swri.org, Southwest Research Institute, CNWRA, 6220 Culebra Rd, San Antonio, TX, 78238, United States
K. S. Chan
Affiliation:
kchan@swri.org, Southwest Research Institute, San Antonio, TX, 78238, United States
D. S. Riha
Affiliation:
driha@swri.org, Southwest Research Institute, San Antonio, TX, 78238, United States
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Abstract

Cladding failure of fuel rods caused by hydride-induced embrittlement is a reliability concern for spent nuclear fuel after extended burnup. Uncertainties in the cladding temperature, cladding stress, oxide layer thickness, and the critical stress value for hydride reorientation preclude an assessment of the cladding failure risk. A set of micromechanical models for treating oxide cracking, blister cracking, delayed hydride cracking, and cladding fracture was developed and incorporated in a computer model. Results obtained from the model calculations indicate that at temperatures below a critical temperature of 318.5 °C [605.3 °F], the time to failure by delayed hydride cracking in Zr-2.5%Nb decreased with increasing cladding temperature. The overall goal of this project is to develop a probabilistic-micromechanical methodology for assessing the probability of hydride-induced failure in Zircaloy cladding and thereby establish performance criteria.

Keywords

embrittlementZrwaste management

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References

REFERENCES

1. Marshall, R.P. and Louthan, M.R., Transactions of ASM 63, 693–700 (1963).Google Scholar
2. Northwood, D.O. and Kosasih, U., International Metals Review 28, 92–121 (1983).Google Scholar
3. Chan, K.S., Journal of Nuclear Materials 227, 220–236 (1996).Google Scholar
4. Einziger, R.E. and Kohli, R., Nuclear Technology 67, 107–123 (1984).Google Scholar
5. Pescatore, C., Cowgill, M.G., and Sullivan, T.M., “Zircaloy Cladding Performance Under Spent Fuel Disposal Conditions,” BNL52235. Upton, New York: Brookhaven National Laboratory. 1989.Google Scholar
6. Raju, I.S. and Newman, J.C., Journal of Pressure Vessel Technology. 104, 293–298 (1982).Google Scholar
7. Newman, J.C. and Raju, I.S., Journal of Pressure Vessel Technology 102, 342–346 (1980).Google Scholar
8. Murakami, Y., “Stress Intensity Factors Handbook,” Vol.2 (Pergamon Press, 1987) pp.751–758.Google Scholar
9. Shi, S.-Q. and Puls, M.P., Journal of Nuclear Materials 208, 232–242 (1994).Google Scholar
10. Simpson, L.A. and Puls, M.P., Metallurgical Transactions 10A, 1093–1105 (1979).Google Scholar
11. Sagat, S. and Puls, M.P. in 17th International Conference on Structural Mechanics in Reactor Technology (SMiRT 17), (Prague, Czech Republic, 2003) Paper No. G06–4.Google Scholar