The corrosion behavior of Zr alloys depends on the kind, size, and distribution of the intermetallic second-phase particles. TEM examinations of Zr-Sn-Fe-Cr alloys irradiated in PWRs at temperatures between 300 and 370°C and fast fluences in the range of 5E21 to 1.3E22 cm-2 have been performed to study the irradiation-induced effects on the precipitates. The alloys contained different types of second-phase particles such as Zr(Fe,Cr)2, Zr2(Fe,Si), and Zr3Fe before irradiation. The influence of irradiation was found to depend on temperature and type of second-phase particles. At temperatures below 310°C, the Laves phase Zr(Fe,Cr)2, which normally is the most frequent precipitate in Zry-4, depletes in Fe, becomes amorphous, and dissolves completely at higher fluences. With increasing temperature, the rate of Fe depletion and dissolution decreases and new Zr-Fe phases are formed. At temperatures above 370°C, the Laves phase remains stable or even grows under irradiation. In Fe-containing Zr alloys with little or no Cr, rather large Zr3Fe precipitates are the most frequent particles. These particles are not dissolved by irradiation even at low temperatures. This was confirmed by annealing after irradiation. As a hypothesis, it was assumed that the different behavior of the various precipitates can be related to their melting or decomposition temperatures by using the homologous temperature (i.e., the temperature under consideration in K normalized to the melting or decomposition temperature in K). This interrelationship has been found to apply for irradiation-induced amorphization. The empirical approach to describe the thermal ripening behavior of second-phase particles before irradiation and to describe the transition from irradiation-induced dissolution to irradiation-induced growth by a normalized (homologous) temperature led to reasonable results.

No abstract

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.

No abstract

Corrosion and dimensional behavior of Zr-Sn-Nb-FeCrV alloys with varying conditions and compositions having been tested out-of-pile and after irradiation in “hot” PWRs for long exposure times (up to 9 annual cycles) and very high burnups (98 MWd/kgU). The exploratory program for alternative zirconium-based alloys performed on fuel rod cladding and corrosion coupons allowed an understanding of the separate effects of composition and fabrication and yielded Zr alloys appropriate for high-burnup, high-fuel-duty application. In the alloy system Zr-Sn-FeCrV, the corrosion rate decreases with decreasing Sn and increasing Fe+Cr+V content, much more pronounced in-PWR than out-of-pile. The addition of Nb exhibits a similar effect as Fe+Cr+V, but increases corrosion in general if Sn > 0.4%. Zr-Sn-FeCr alloys are sensitive to the thermal treatment during fabrication and therefore have to be fabricated according to a well-defined A-parameter to achieve high corrosion resistance. Nb and V containing Zr-Sn-Fe alloys allow fabrication in a larger A-parameter range. The beneficial effect of high cold work and a low degree of recrystallization on corrosion is significantly reduced if the Sn content is low. At high fluences, some alloys exhibit an increase in the corrosion rate. With decreasing Sn content, the onset of the acceleration is shifted towards higher fluences and its extent is reduced. Alloys containing Sn < 0.2% do not exhibit increased corrosion induced by high fluences, at least up to 1.6E22 cm-2, E > 1 MeV. Irradiation induced and thermal creep depend on the content of Sn, Nb, and O in solid solution, on the grain size, and on the degree of recrystallization. Growth of stress-relieved and partially recrystallized fuel rod claddings strongly depends on the Sn content: it is low if Sn < 0.5%. In addition, the C content and the grain size affect growth of fuel rods.