Pellet-clad interaction

Material problems in a nuclear reactor plant can be grouped into at least two categories, one concerning the nuclear reactor core and one that will apply to all plant materials. This chapter discusses specific material problems associated with the reactor that include pellet-cladding interaction, fuel densification, fuelcladding embrittlement, and effects on fuel due to inclusion and core burnup. 

FIG. 7.29. Pellet cladding interaction in a FFTF test absorber pin. 

Assuming that pellet-clad interaction should happen to occur with high bum-up fuel, there is a possibility of the cladding being damaged due to the stresses that would be created by a too rapid increase in the reactor power. This effect was in fact observed in a small number of cases in the mid-1970s and, as a result the rate of power increase after a refueling shutdown in PWRs is now subject to a limitation of about 3% increase per hour. Based on experience accumulated over the past few years, this restriction would appear... 

To summarize, the suitability of UO2 as a reactor fuel for both thermal and fast reactors has been confirmed by a wealth of experience, and the failure rate of fuel pins can be made acceptably low (down to about 1 in 10" ) by careful quality control in manufacturing. Particular problems which have arisen, such as fuel densification and damage due to pellet-clad interaction, have been solved by adjustment of the initial fuel density and pin gas pressure, and by limiting the rate at which large power increases are allowed to occur. Alternatives to the oxide fuel, such as uranium carbide and uranium nitride, have been investigated less extensively, but hold promise of superior performance in fast reactor systems. 

Deeper knowledge is required on the properties of the fuel composition, the fuel pellet-cladding interaction (FPCI), which is the basis for the RBEC-M core performance. 

Wood, J.C., Surette, B.A., Aitchison, I., Clendening, W.R., 1980. Pellet cladding interaction evaluation of lubrication by graphite. Journal of Nuclear Materials 88 (1), 81—94. 

Pellet—cladding interaction has not been a major focus of the on-going SCWR research as, for the most part, data under relevant conditions already exist for similar fuel and fuel cladding materials. Oka et al. [3] note that while the thermal expansion coefficient of the Zr alloy cladding used in LWRs is smaller than that of the UO2 fuel pellets, the thermal expansion coefficients of the stainless steels or nickel-base alloys being considered for the SCWR are expected to be close to that of UO2. Pellet—cladding interaction will need to be considered during future SCWR fuel qualifications tests. 

Similarly, it is commonly known that the centerline melting criterion is conservative. The pellet centerline melting may lead to excessive fuel volume expansions or FP gas releases, which may cause pellet cladding interaction (PCI) or excessive... 

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