SchlieBlich erhiilt man aus G1. (30) unter Beriicksichtigung von G1. (32) den Schadigungsmodul GS gemaB G1. (23a) und fiir t + 0 den Ursprungsmodul G der Klebschicht in Abhangigkeit von der Schichtdicke D zu [2] (34) G1. (34) ist in den Abb. 17 und 18 experimentellen Ergebnissen gegenubergestellt.

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.

There is growing interest among nuclear power utilities to increase power output and to maximize electricity production. In order to deal with these more challenging conditions, new alloys have been developed. Another approach is to apply protective coatings on existing cladding zirconium alloys. Although the coatings described herein were developed in order to improve the corrosion behavior under normal conditions, it is clear that these coatings can also be applied to other materials (iron-chromium-aluminum, molybdenum, silicon carbide) presently being studied for accident-tolerant fuel. In 2011, several commercially available coatings applied by physical vapor deposition were proposed to improve the performance of existing cladding materials. Since then, these coatings have been investigated in out-pile and in-pile experiments at the Halden Reactor Project from 2011 to 2014. Boiling and pressurized water reactor (PWR) in-pile experiments were performed on small cylindrical Inconel 600 and Zircaloy-4 samples with CrN, TiAlN, and ZrO2 coatings. In all cases examined, the CrN coating came out as superior and remained completely intact after the irradiation, as evidenced by SEM analysis. Based on these promising results, an in-pile experiment with coated Zircaloy claddings (containing fuel) was performed in the Halden reactor in PWR conditions in 2014. The experimental rig contained four Zircaloy-4-clad fuel rods—three coated (CrN, TiAlN, and AlCrN) and one uncoated. Due to a mechanical deformation of tubes for cooling water within the rig, the fuel rods were insufficiently cooled, resulting in a higher-than-normal cladding temperature. Nevertheless, an SEM examination showed that the CrN coating largely remained intact apart from several local spots where the coating had cracked or disappeared. Underneath the cracked coating, oxide formation was observed in the Zircaloy-4 cladding. The TiAlN and AlCrN coatings both disappeared after the irradiation.

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