Zircaloy-2 fuel cladding is susceptible to several forms of secondary degradation following steam ingress in defective rods in BWRs. Hydride blister, circumferential, and axial cracks have been reported. An in-core test was performed at the Halden reactor to evaluate the root cause of the secondary degradation, with particular attention paid to the axial cracking issue. The test included four instrumented rods of one meter long. Three rods had Zr-lined Zircaloy-2 cladding, and the other was a nonlined cladding. Two initial diametral gap sizes and two steam ingress conditions were designed. The rods were irradiated to 8 GWd/MtU before steam was admitted into the rods when the rods were operated at power. Subsequently, a power ramp from 33 to 43 kW/m was executed to cause mechanical stresses on the cladding. The test was terminated following detection of a significant activity release. Three of the four rods, including two Zr-lined and the nonlined ones, exhibited hydride bulges and short axial cracks; the remaining Zr-lined rod did not develop secondary defects. The secondary defects developed first by heavy localized hydriding in all cases. Steam ingress causes increase in fuel temperature, formation of central voids, and pellet swelling at the initial hydride defects, resulting in local plastic strains on the cladding at the crack tip. Susceptibility of the cladding to cracking depended on the hydriding characteristics of the cladding. Unlined Zircaloy cladding was susceptible to sunburst hydriding locally within a short axial length. Zr-lined cladding was more resistant to localized sunburst hydriding; hydrogen was absorbed more uniformly by the cladding over a substantially longer cladding length and formed thick hydride rims near the cladding outer surface. The hydride rim ahead of the crack tip in a Zr-lined cladding produced radial hydrides intruding deep into the cladding wall under plastic strains. Repeated crack tip straining by central void formation and deep radial hydride penetration into the cladding from hydride rims would assist the crack to penetrate the cladding wall perpendicular to the cladding surfaces and propagate axially over a long distance. Unlined Zircaloy cladding can also propagate axial cracks by the same mechanism, but its susceptibility to localized sunburst hydriding would limit the length of such axial cracks to a short length nearby the sunburst hydride region. Results from this program suggest that the hydriding characteristics of the cladding inner surface determine the susceptibility of the cladding to long axial cracking. A low corrosion resistance Zr-liner and, particularly, a dry hydrogen environment, which accelerates the hydriding rate, can increase the rate of axial cracking. The findings from this program are consistent with observations of high susceptibility of high purity Zr-liner to axial cracking in BWRs, and provide basis for its mitigation.

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A plutonia stabilised zirconia doped with yttria and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er)yPuxZr1-yO2-ζ where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O2-ζ (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia-IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO2. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O2 fuels. The properties of the spent fuel pellets are presented focusing on the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO2 in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a burn and bury strategy.