Friedrich Garzarolli

Behavior and Properties of Zircaloys in Power Reactors: A Short Review of Pertinent Aspects in LWR Fuel

Zircaloy-2 and -4, developed mainly in the US, have been used in Germany for fuel rod claddings and in-core structural components from the beginning of reactor technology. Extensive studies of the material properties of the Zircaloys have been performed in Siemens laboratories since 1957. Zircaloy-2 and -4 turned out to be very reliable materials that fulfilled all requirements for normal operation and likewise the requirements for postulated accidental conditions and for intermediate storage for many years. Optimization of Zircaloy-2 and -4 during recent years includes both optimization of microstructure and of chemical composition. BWRs and PWRs need differently optimized materials. Today`s more demanding operation conditions and discharge burnups required a further optimization of the Zircaloys and for hot PWRs even the development of more corrosion-resistant Zr alloys. A significant improvement of PWR corrosion behavior can be achieved with Zr alloys using the alloying elements of Zircaloy with somewhat modified concentrations. Sn should be below or at least in the lower range of the ASTM specification range for Zircaloy-4, Fe and Cr should be somewhat higher, and Si should be specified as an alloying element rather than as an impurity.

Microstructure and Corrosion Studies for Optimized PWR and BWR Zircaloy Cladding

The increased discharge burnups envisaged today require an optimization of the Zircaloy cladding properties with regard to corrosion with different strategies for optimizing the corrosion for boiling water reactor (BWR) and pressure water reactor (PWR) applications. A broad laboratory study showed that the influence of temperature and environmental details upon autoclave test results is rather complex. Zircaloy with fine (∼0.05 μm), uniformly distributed intermetallic precipitates reveals nodular corrosion in steam at 420°C or below and excessive uniform corrosion in water at 350°C, but the best nodular behavior in steam at 500°C. However, samples with larger intermetallic precipitates (>0.2 μm) exhibit a strong tendency to nodular corrosion in 500°C steam, but good behavior in steam at 420°C or below and in water at 350°C. The tendency toward nodular corrosion in 400 to 500°C steam can be suppressed by additions of oxygen; the ranking in uniform corrosion in 350°C water can be altered by additions of hydrogen. The latter was observed on samples exposed up to 1000 days in degassed water and afterwards 1000 days in water with a hydrogen overpressure. Similar correlations were found for in-pile exposure with some exceptions. Materials with very fine intermetallic precipitates, which show little tendency for nodular corrosion and behave well in-pile in oxygenated (mostly BWR) environments, exhibit accelerated uniform corrosion in a hydrogenated (PWR) environment with a corrosion rate that depends very little on temperature. However, in oxygenated coolant, materials with coarse precipitates indicate severe nodular corrosion, which is almost independent of the temperatures within the range of 290 and 350°C. As a result of these findings a dual strategy is technologically feasible by using different values of the accumulated annealing parameters (Σ A,) for BWR and PWR application. For optimized BWR cladding tubes Σ A, ≤ 10 - 1 8 h is proposed; the best range for PWR cladding appears to be at 2 x 10 - 1 8 h ≤ Σ A i ≤ 5 x 10 - 1 h.

Effects of High Neutron Fluences on Microstructure and Growth of Zircaloy-4

Irradiation of Zircaloy affects its microstructure and macroscopical properties, for example, influencing its irradiation growth. To gain more insight into these phenomena, experimental fuel rods and growth specimens withvarious fabrication parameters were irradiated in a pressurized water reactor (PWR) to high fluences. Some of the growth specimens were exposed to a fast neutron fluence of up to 2.3 x 10 2 2 cm - 2 (≥0.82 MeV) over a period of 10 years. Following exposure, the irradiation-induced alterations of the microstructure and the intermetallic precipitates were studied by optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM). At a temperature of 300°C during irradiation to fluences up to 7 x 10 2 1 cm - 2 , growth increases with increasing yield strength. Recrystallized material, which has a low yield strength, exhibits an increased growth rate at very high fluences (≥1 x 10 2 2 cm - 2 ). Postirradiation annealing studies indicate that the early irradiation growth of the recrystallized material can be recovered, whereas the later accelerated growth does not seem to be recoverable. At temperatures in the range of 330 to 350°C, the growth depends on the grain size, especially below 2 μm, and on the carbon-content. The effect of yield strength on growth was less at 330 to 350°C than at 300°C, probably because of an irradiation-induced recovery. Moreover, TEM showed that an irradiation-induced formation of dislocations with a c-component occurs at neutron fluences ≥9 × 10 2 1 cm - 2 , whereas dislocation loops or fine precipitates or both form at lower fluences. The intermetallic precipitates observed in the microstructure of the unirradiated, initial material exhibit two types of intermetallics. One type contains (Fe + Cr + Zr) and the other type contains only (Fe + Zr). The effect of irradiation on those intermetallics depends on temperature. At ≤300°C the (Fe + Zr)-type intermetallic precipitate dissolves at neutron fluences above 5 x 10 2 1 cm - 2 . The (Fe + Cr + Zr)-type precipitates become more and more amorphous and release iron to the matrix resulting in a decreasing Fe/Cr ratio. The diameter and the number of the precipitates decrease with increasing neutron fluences at this temperature. Only a few small precipitates can still be observed after a neutron fluence of 1.5 x 10 2 2 cm - 2 . At temperatures above 340°C the size of intermetallics increases because of irradiation enhanced ripening.

Spannungssimulation an hüll werkstoffen bei rohrförmigen brennelementen

Zusammenfassung Die Entwicklung eines integrierten Siede-Uberhitzerreaktors fuhrte zum Konzept eines rohrformigen Brennstabes mit axialen mechanischen Belastungen der Hullrohre, die mit dem Lastzustand des Reaktors wechseln. Zur experimentellen Prufung des theoretischen Spannungsmodells fur dieses Brennelement wurde eine Spannungs-Simulationseinrichtung entwickelt, die es gestattet, an Hullrohrabschnitten mit realistischen Durchmessern und Wandstarken die Spannungen und die plastischen Verformungen in den Brennstabhullen quantitativ in Abhangigkeit von der Betriebstemperatur zu verfolgen. Die bisher durchgefuhrten Messungen bestatigen das theoretische Modell. Daruberhinaus wurden erste Messwerte uber den Betrag der plastischen Wechselverformung gewonnen. Die Versuche werden fortgesetzt mit dem Ziel, die fur die Auslegung der Brennstoffstabe wichtigen Materialkennwerte zu ermitteln.