Laboratory corrosion tests have always been an important tool for Zr alloy development and optimization. However, it must be known whether a test is representative for the application in-reactor. To shed more light on this question, coupons of several Zr alloys were exposed under isothermal conditions in all or most of the following environments: In-Reactor: (1) PWR core at 300 to 340°C up to six years. (2) BWR core with a low sensitivity to nodular corrosion up to four years. (3) BWR core with a high sensitivity to nodular corrosion up to two years. Ex-Reactor (in Autoclave): (1) 350°C/pressurized water up to three years. (2) 400°C/100-bar steam up to two years. (3) 350°C/0.01 M LiOH water up to two years. (4) 500 to 515°C/high-pressure steam 16 to 24 h. In addition, the material condition of several of the examined Zr alloys was varied over a wide range. For evaluation of the in-PWR tests and for comparison of out-of-pile and in-pile tests, the different temperatures and times were normalized to a temperature-independent “normalized time” by assuming an activation temperature (Q/R) of 14 200 K. Comparison of in-PWR and out-of-pile corrosion behavior of Zircaloy shows that corrosion deviates to higher values in PWR if a weight gain of about 50 mg/dm2 is exceeded. In the case of the Zr2.5Nb alloy, a slight deviation of corrosion as compared to laboratory results starts in PWR only above a weight gain of 100 mg/dm2. In BWR, corrosion of Zircaloy is enhanced early in time if compared with out-of-pile. Zr2.5Nb exhibits higher corrosion results in BWR than Zircaloy-4. Alloying chemistry and material condition affect corrosion of Zr alloys. However, several of the material parameters have shown a different ranking in the different environments. Nevertheless, several material parameters influencing in-reactor corrosion like the second phase particle (SPP) size or in-PWR behavior as the Sn and Fe content can be optimized by out-of-pile corrosion tests.
The corrosion of Zircaloy-4 is accelerated in water or steam above about 350°C when LiOH is present in sufficient concentrations. A short-term test (16 h) to predict inreactor corrosion behavior has been developed based on this accelerated corrosion. A group of claddings with a range of in-reactor corrosion behavior has been tested in LiOH solutions from 310 to 415°C with Li concentrations up to 375 ppm. Up to 100 times the rate seen in pure water was obtained in supercritical water with high LiOH levels. The best correlation to the in-reactor behavior was obtained for a Li content in the range of 75 to 190 ppm in the temperature range of 390 to 410°C. The corrosion rate for Zircaloy-4 in steam containing LiOH has been observed to be higher than in water with LiOH for tests below the critical temperature. For tests above the critical temperature, the corrosion rate was greater for samples which were above the initial water line in the capsule. These samples would have been exposed to LiOH-bearing steam during the initial heat up of the samples. Therefore sufficient Li is present in the steam phase to cause accelerated corrosion. The accelerated test with LiOH correlated well with 415°C/3 day autoclave test results for Zircaloy-4 but not for zirconium-based alloys containing niobium. The 415°C/3 day test has been found to correlate well to in-reactor corrosion performance for Zircaloy-4. Based on the assumption that this test is also effective for evaluating the corrosion resistance of other zirconium-based alloys, the accelerated LiOH test appears to be effective only within the Zircaloy-4 composition range. Therefore the LiOH accelerated test should not be used for evaluating new alloy compositions.
The effects of variations in the beta-quench parameters during the tube reduction process on the final microstructure and microchemistry of Zircaloy-2 and Zircaloy-4 cladding were studied by TEM/STEM and correlated with corrosion rates in 400 and 500°C steam autoclaves. The Process A Zircaloy-2 specimens (recrystallization annealed) revealed primarily two types of ternary intermetallic compounds: tetragonal Zr2(Fe,Ni) and hexagonal Zr(Fe,Cr)2. The solutelean Zr2(Fe,Ni) particles were much coarser than the solute-rich Zr(Fe,Cr)2 particles. The Process A Zircaloy-4 specimens revealed only one type of ternary particles, Zr(Fe,Cr)2, but with much higher Fe/Cr ratios and mean particle size than those of the corresponding particles in Zircaloy-2. Also, both the cubic and hexagonal forms of Zr(Fe,Cr)2 were noted in this sample. More aggressive beta quenching reduced the mean particle size and volume fraction of the intermetallic particles and increased their number density in both alloys; however, the changes were greater in Zircaloy-2. In both alloys, the degree of these changes was most pronounced when beta quenching was performed close to the final tube reduction step. In Zircaloy-2, the Zr-Fe-Cr type particles showed higher refinement than the Zr-Fe-Ni type particles. The variations in Zircaloy microstructure with the timing and aggressiveness of beta quenching in the tube reduction process were attributed to differences in quench severity and thermal exposure after the quench. A critical discussion of the influence of alloy chemistry was also presented to reconcile the differences in beta-quenching response between Zircaloys 2 and 4. The 500°C steam autoclave corrosion data revealed significant improvement when beta quenching was performed close to the final tube dimensions.
Milestones Porous ceramic substrate materials were evaluated in order to obtain a ceramic substrate material to replace the glass substrates. A ceramic substrate material was selected for use that is superior to any material that has been tried previously. Metallizing techniques have been further improved and a better understanding of the techniques has been obtained. Physical parometers of the substrates were recorded and.
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