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.

Two major factors that affect corrosion behavior of Zircaloy-type alloys have been evaluated in a PWR cladding development program ongoing since the early eighties. One of the important parameters was found to be the alloying concentration and distribution of transition elements. The influence of transition elements (Fe, Cr, and V) was studied on corrosion coupons and experimental fuel rods. Isothermal long-time corrosion tests were performed out-of-pile (1 to 3 years in 300 to 370°C) pressurized water and 400°C steam with and without LiOH additions) and in PWRs (up to 6 years at 310 to 335°C). Experimental fuel rods were irradiated in PWRs up to nine annual cycles. The experiments show a large reduction of corrosion rate with increasing content of transition elements in all environments. Above a certain transition element content no further improvement in corrosion is observed. The test results clearly suggest the use of transition element concentrations higher than specified by ASTM for Zircaloy-2 and -4. The beneficial effect of transition elements increases with increasing exposure time. The most significant effect of corrosion behavior was seen after six years of PWR exposure, especially at high temperatures. Different transition elements influence corrosion differently, depending on the environment. Out-of-pile in 350 to 370°C pressurized water and 400°C steam, a beneficial effect is exhibited only for Fe, but not for Cr and V. In PWR and in diluted LiOH autoclave tests, all three transition elements reduce corrosion. for Zr-based alloys with transition element (TE) concentrations meeting the ASTM Zircaloy-4 specification range, the fraction of corrosion hydrogen picked up by the metal increases with increasing corrosion resistance. Zr-based alloys with enhanced transition element concentrations (above the ASTM limits) exhibit reduced hydrogen pickup fractions.