The influence of the composition of Zircaloy-type alloys has been evaluated previously with respect to mechanical properties (1973 International Conference Zirconium in the Nuclear Industry, ASTM STP 551) and to nodular corrosion (1985 International ASTM Conference on Zirconium in the Nuclear Industry, ASTM STP 939). The present study extends this previous work to include uniform corrosion, which is of interest for cladding tubes and spacer grids of fuel assemblies. A variety of ingots with compositions both within and outside of the ASTM specification range for Zircaloy-2 and Zircaloy-4 have been tested. The material was fabricated into strips using a conventional fabrication procedure, which incorporated an intermediate beta quench followed by working and annealing in the upper alpha-range. The alloy elements covered the following ranges: tin: 0.2 to 1.7%; iron: 0.05 to 0.53%; chromium: 0.04 to 1.05%; and nickel: 0.003 to 0.046%. In addition, oxygen, carbon, silicon, and phosphorus were varied over the range of standard Zircaloy contents. Uniform corrosion was studied out of pile by long-time autoclave testing in pressurized water at 350°C (380 to 840 days) and in high pressure steam at 400°C (180 to 397 days). These data supplement testing previously reported for these same alloys in steam at 500°C under static and refreshing conditions (Strasbourg, 1985, ASTM STP 939). Electron microscopic examinations (SEM and TEM) have been used to characterize the intermetallic precipitates to clarify the effect of the various elements. The results of the 350 and 400°C testing show that the time to transition from the cubic to linear rate increases and the post transition rate decreases with decreasing tin and carbon and increasing silicon content. Oxygen and phosphorus have not shown an influence in the range studied. The effect of iron and chromium was more complex than for these other elements; the effect was also different in nature between water and steam tests. Markedly accelerated corrosion was seen at Fe + Cr concentrations below the limits of the Zircaloys (⩽0.15%) especially in steam. This composition effect is similar to that previously reported for the same alloys tested at 500°C. High Cr/Fe ratios often resulted in high corrosion even at moderately high levels of both elements. The effect of iron and chromium can be correlated to type, size, and frequency of the intermetallics.
The performance of Zr1NbSnFe alloys within the range of Sn 0–0.65 % and Fe 0.03–0.35 % were studied through irradiation of fuel rods in two pressurized water reactors (PWRs) operating with significantly different fuel management strategies. Material test rod irradiations have also been launched in order to determine irradiation growth and corrosion behaviour on tubes irradiated under conditions representative of guide tubes. Results show that the increase in tin content up to 0.3 % does not significantly change the corrosion resistance nor the hydrogen pick-up compared to Zr1Nb alloy, while ensuring a higher creep resistance and an improved dimensional stability. On the contrary, at 0.5 % Sn, the corrosion resistance can be significantly degraded under demanding conditions. The iron addition to the alloy can be considered as a second order parameter for both corrosion and creep properties.
The data base on the corrosion behavior of Zr alloy materials under BWR conditions was evaluated with respect to the burnup target of 70 MWd/kgU. At high burnups, corrosion rate and the rate of hydrogen pickup (HPU) may increase. This onset of increase obviously depends on the material, but also seems to be significantly affected by the coolant water chemistry. Because small differences in corrosion behavior at lower burnup might become more and more important with increasing burnup, Framatome ANP has performed several studies on the separate and combined effects of (1) alloying content of the claddings, (2) cladding material condition, (3) impurity content of the cladding, and (4) the coolant chemistry. This paper focuses on the effects the concentration of alloying elements and of impurities (including microstructural differences imposed by the annealing treatment) have on corrosion. The corrosion effects were evaluated in material test irradiation programs in two BWRs. Zircaloy type materials processed at low temperatures (LTP), defined by a low particle growth parameter (PGP) value, exhibit a maximum corrosion resistance between 1.2 and 1.5 % Sn. Impurities, such as C, O, and P can increase the corrosion of Zircaloy in BWRs at high burnup. The higher the corrosion resistance of the base material, the more pronounced is the increase seen at high burnup. Above a critical PGP value, in-pile corrosion increases. At high burnups, Zry-4 shows a higher increase with increasing PGP than Zry-2, whereas at lower burnups both behave similarly. The critical PGP value varies with the chemical composition, such as Fe, Cr, and Ni content and the distribution of second phase particles (SPP). The effect of Si is more complex. Si increases in-pile corrosion at contents in excess of 140 ppm. Contents at 80 to 140 ppm can be beneficial, when the β-quench rate applied during fabrication is not high enough to ensure a uniform distribution of the SPP, and the alloying composition and the concentration of impurities is in a beneficial range. The hydrogen pickup fraction (HPUF) of Zircaloy type samples in BWRs decreases with decreasing corrosion resistance but differs from plant to plant. There are indications that the difference can partially be attributed to the Fe content in the coolant. The results are in agreement with the irradiation experience with Zry-2 LTP cladding extending up to 73 MWd/kgU in different BWRs.
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