Effect of Absorbed Hydrogen on the Stress Corrosion Cracking (SCC) Susceptibility of Unirradiated Zircaloy Cladding

Effect of absorbed hydrogen on the stress corrosion cracking (SCC) susceptibility of unirradiated Zircaloy cladding was examined. The data obtained from literatures showed that the normalized ratios of SCC threshold stress (σ_th ) to 0.2% yield stress (σ_0.2.) claddings, from which the influence of σ_0.2 had been eliminated, increased with increasing hydrogen contents below 50ppm in unirradiated Zircaloy-2 and -4. For Zircaloy-4, the break point was observed in the relationship between normalized ratios of σ_th to σ_0.2 and hydrogen content in sample at hydrogen content of approximately 50ppm. Thermodynamic calculations were carried out for the reaction between iodine gas and zirconium containing hydrogen. The results suggested that the reactions hardly occurred at increased hydrogen content and zirconium reacted with iodine gas only below 100 ppm of hydrogen. Since these tendencies correspond to those of the normalized ratios of σ _th to σ_0.2 on the hydrogen content, it is considered that hydrogen affects the reactions between iodine gas and zirconium and reduces the SCC susceptibility of Zircaloy cladding.     

Comparative study and imaging by PhotoElectroChemical techniques of oxide films thermally grown on zirconium and Zircaloy-4

The oxidation of iron and chromium that are present as impurities in zirconium metal or as alloying elements in Zircaloy-4 was investigated with PhotoElectroChemical techniques (PEC), highlighting the chemical nature, the size and the lateral distribution of Fe and Cr-containing phases in thin zirconia scales formed during the oxidation of pure zirconium and Zircaloy-4 at 470 °C in oxygen. In the case of zirconium, iron and chromium impurities led to the formation of oxides distributed in a homogeneous way in the zirconia scale, while in the case of Zircaloy-4 these elements, present in the form of intermetallic particles in the substrate, led to the formation of localised haematite Fe₂O₃, rhomboedric solid solution (Fe_xCr_1−x)₂O₃ and chromia Cr₂O₃ phases. These phases were accurately studied via the measurement of their respective band-gap (Fe₂O₃: 2.2 eV, (Fe_xCr_1−x)₂O₃: 2.6 eV and Cr₂O₃: 3.0 eV). It is concluded that PEC techniques represent a sensitive and powerful way to locally analyse the various semiconductor phases in the oxide scale at a micron scale.     

Evaluation of SCC crack behavior in zirconium and zircaloy-2 using nonlinear fracture mechanics parameters

Applicability of nonlinear fracture mechanics parameters, i.e. J-integral, crack tip opening displacement (CTOD), and crack tip opening angle (CTOA), to evaluation of stress corrosion crack (SCC) propagation rate was investigated using fully annealed zirconium plates and Zircaloy-2 tubing, both of which produce SCC with comparatively large plastic strain in an iodine environment at high temperatures.Tensile SCC tests were carried out at 300°C for center-notched zirconium plates and internal gas pressurization SCC tests at 350°C, for Zircaloy-2 tubing, to measure the SCC crack propagation rate. The J-integral around semi-elliptical SCC cracks produced in Zircaloy-2 tubing was calculated by a three-dimensional finite element method (FEM) code.The test results revealed that the SCC crack propagation rate dc/dt could be expressed as a function of the J-integral, which is the most frequently used parameter in nonlinear fracture mechanics, by the equation dc/dt = C · Jⁿ, where C and n were experimental constants.Among the other parameters, CTOD and CTOA, the latter appeared to be useful for assessing the crack propagation rate, because it had a tendency to hold a constant value at various crack depths.     

The effect of alloying modifications on hydrogen uptake of zirconium-alloy welding specimens during corrosion tests

The hydrogen uptake behavior during corrosion tests for electron beam welding specimens made out of Zircaloy-4 and zirconium alloys with different compositions was investigated. Results showed that the hydrogen uptake in the specimens after corrosion tests increased with increasing Cr content in the molten zone. This indicated that Cr element significantly affected the hydrogen uptake behavior. Fe and Cr have a low solubility in α-Zr and exist mainly in the form of Zr(Fe,Cr)₂ precipitates, which is extremely reactive with hydrogen in its metallic state. It is concluded that the presence of Zr(Fe,Cr)₂ second phase particles (SPPs) is responsible for the increase in the amount of hydrogen uptake in the molten zone of the welding samples after corrosion, as Zr(Fe,Cr)₂ SPPs embedded in α-Zr matrix and exposed at the metal/oxide interface could act as a preferred path for hydrogen uptake.     

Effects of ECC Water Injection Rate on Reflood Phase during LOCA

Some metal iodides such as of Fe, Al, Zr and Te are known to cause stress corrosion cracking (SCC) of Zircaloy just as iodine itself does. Therefore 15 metal iodides were selected as corrodants, and SCC tests were carried out using the internal gas pressurization method.

The results showed that: (1) only those metal iodides which react thermodynamically with Zr to produce ZrI₄ cause SCC of Zircaloy-2; (2) when SCC occurs, the reaction rate between the iodide and Zr seems to be a main factor in determining the SCC susceptibility; (3) gaseous ZrI₄ is the most corrosive agent; and (4) some species of metal iodides, such as PbI, cause SCC of Zircaloy-2 more easily than I₂ vapor.

Scanning electron microscope (SEM) examination and electron probe microanalysis (EPMA) on the fracture surface of failed specimens revealed that ZrI₄, formed as the reaction product between the metal iodides and Zr, might induce SCC of Zircaloy-2 rather than the iodides themselves.     

High-temperature oxidation of zircaloy-2 and zircaloy-4 in steam

At temperatures above the $\text{(α + β)}\text{β}$ transformation temperature for zirconium alloys, steam reacts with β-Zr to form a superficial layer of zirconium oxide (ZrO₂) and an intermediate layer of oxygen-stabilized α-Zr. Reaction kinetics and the rate of growth of the combined (ZrO₂ + α-Zr) layer for Zircaloy-2 and Zircaloy-4 oxidation in steam were measured over the temperature range 1050–1850°C. The reaction rates for both alloys were similar, obeyed parabolic kinetics and were not limited by gas phase diffusion. The parabolic rate constants were consistently less than those given by the Baker and Just correlation for zirconium oxidation in steam. A discontinuity was found in the temperature dependence of both the reaction rate and the rate of growth of the combined (ZrO₂ + α-Zr) layer. The discontinuity is attributed to a change in the oxide microstructure at the discontinuity temperature, an observation which is consistent with the zirconium-oxygen phase diagram.     

Chemical aspects of iodine-induced stress corrosion cracking failure of zircaloy-4 tubing above 50°C

The chemical environment associated with iodine-induced SCC failure of Zircaloy-4 tubing above 500°C has been characterized. At the critical iodine concentrations which result in SCC initiation and propagation, most of the iodine is present as condensed zirconium subiodides (I/Zr ? 0.4). Only a small part of the iodine remains in the gas phase as ZrI₄. The gaseous ZrI₄ is probably responsible for crack initiation and propagation. The critical ZrI₄ pressures for SCC failure have been estimated in zircaloy/iodine reaction experiments performed with unstressed zircaloy tube specimens. These pressures were confirmed in additional creep rupture tests conducted under controlled ZrI₄ partial pressure conditions. The estimated critical ZrI₄ pressure above which low-ductility SCC failure of the zircaloy tubing always occurs, independent of time-to-failure, varies between 0.005 bar at 550°C and 0.043 bar at 800°C. Below the critical values, however, a rather wide range of ZrI₄ pressures is associated with the onset of the SCC, especially at temperatures below 800°C. A comparison of the experimental results with available thermochemical data in the Zr-I system indicates that the main reaction involved during crack propagation is chemisorption of iodine-containing species on the fresh zircaloy surfaces created by metal straining at the crack tip.     

The Effect of Oxide Microstructure on Kinetic Transition in Out-of-pile Steam Corrosion Test for Zircaloy-2 and Nb-added Zircaloy-2

In order to study the mechanism of kinetic transition of corrosion rate for zirconium alloys, oxide films formed on Zircaloy-2 (Zry-2) and Nb-added Zircaloy-2 (0.5Nb/Zry-2) in steam at 673 K and 10.3 MPa were examined with TEM and SIMS.

Kinetic transition occurred at almost the same oxide thicknesses for both Zry-2 and 0.5Nb/Zry-2, but the corrosion rate after the transitions were quite different for the two alloys. Zircaloy-2 showed cyclical oxidation, while the weight gain of 0.5Nb/Zry-2 increased linearly.

The morphology and crystal structure were similar for the oxides of the two alloys and both the oxide films still mainly consisted of columnar grains even after the transition. Interface layers which mainly consisted of a-Zr crystallites were observed for both alloys and the oxygen content in the interface layers increased after the transition.

The solute concentrations of Fe, Cr and Ni became higher, accompanying the increase of oxygen concentrations at columnar grain boundaries in the oxide films after the transition for 0.5Nb/Zry-2. It was thought that the properties of grain boundaries of the 0.5Nb/Zry-2 oxide films changed after the transition, and the increase in oxygen diffusivity at grain boundaries caused the linear increase in weight gain.     

Effect of additives on the susceptibility of zirconium alloys to nodular corrosion

A new approach based on a stability analysis of a uniformly growing oxide film was applied to estimate the effect of alloying additives on the susceptibility of zirconium alloys to nodular corrosion. The analytical results agree with available experimental data on effect of Fe and Ni on resistance of Zircaloy-2 and Zircaloy-4 to the growth of nodular oxide.     

Quantitative characterization of crystallographic textures in zirconium alloys

The techniques for determining inverse pole figures and direct pole figures of zirconium alloys by X-ray diffraction are summarized, and their advantages, disadvantages, and limitations are discussed. A critical review is made of the various parameters that have been used to quantify the texture in zirconium alloys. A new series of four quantitative texture numbers F_T, (SD)_T, F_A, and S, which are obtained from the direct pole figure, are proposed. Pole figures are determined for Zircaloy-2 tubing produced by three tubing manufacturers. The four texture numbers are calculated and are used to compare the textures of the three manufacturers and the through wall texture gradient of one manufacturer.     

Localised electron transport in corroding zirconium alloys

Zircaloy-2, in the as-received and heat-treated conditions, and the binaries of Zr with Sn, Fe, Cr and Ni, were oxidised in a fused salt medium at 300 and 400° C. The electrochemical polarisation behaviour was studied at various oxide film thicknesses. Metal electrodes were evaporated on to the oxide films and the dc conduction of the alloy-oxide-metal diodes was investigated. Analysis of the data shows that the presence of the Zr-Fe intermetallic phase is associated with an enhanced localised electron transport which leads to low negative corrosion potentials and fast oxidation rates; the high temperature oxidation behaviour of zirconium alloys falls into two distinct groups depending on the presence or absence of this intermetallic phase. In proposing a model for the oxidation of zirconium alloys, information regarding the processes controlling ion and electron transport is essential. Because of the possibility of electron transport occurring through the bulk oxide and at localised second-phase precipitates by different mechanisms, it is unlikely that there is a simple relation between the electron current and potential drop across the oxide.     

Mössbauer spectroscopy of tin in unirradiated and neutron irradiated Zircaloys

Mössbauer spectroscopy of the 23.9 keV γ-rays in ¹¹⁹Sn nuclei was applied to study Zircaloy-2, Zircaloy-4, and other tin-bearing zirconium-based alloys of interest to nuclear power technology. Zircaloys are extensively used in nuclear reactors as fuel cladding. In CANDU reactors, Zircaloys are also used as major structural components such as calandria tubes, and were used until the late 1970's as pressure tubes (now replaced by Zr–2.5Nb alloy). Unirradiated specimens of these alloys, as well as radioactive specimens, both neutron-irradiated in high-flux test reactors and extracted from nuclear power-reactor components after many years of service, were examined. The obtained spectra consistently showed tin in substitutional solid solution in α-Zr, whereas no evidence was found of metallic Sn or intermetallic Zr₄Sn precipitates. In oxide scrapes removed from Zircaloy-2 pressure tube of one of CANDU reactors, where the alloy was exposed for about 10 years to pressurized heavy water coolant at temperatures of ∼280°C, a considerable fraction of tin was found in the Sn(IV) state, in the form that coincides with the state of tin in stannic oxide, SnO₂. The same form of tin was identified in filterable deposits in the primary heavy water coolant of CANDU reactors. For comparison, in Zircaloy heated in air, SnO₂ was formed only at temperatures above 500°C.     

Effect of Neutron Irradiation on Iodine Stress Corrosion Cracking Susceptibility of Zircaloy-2 Tubing

The effect of neutron irradiation on the iodine stress corrosion cracking (SCC) of Zircaloy-2 tubing of 8×8 type design was studied by means of ring tension test, using specimens unirradiated and irradiated to 3.2×l0¹⁹ and 3.0×10²⁰ n/cm² (E>lMeV). The SCC threshold stresses were determined from constant load tests and the SCC initiation stresses were defined from constant cross-head rate tests. Both stresses increased with increasing neutron fluence, reaching a maximum at a neutron fluence between 10¹⁹ and 10²⁰ n/cm² and then decreased. The tendency is qualitatively in good agreement with the hypothetical conclusion derived by Lunde & Videm, for SCC failure stresses from internal gas pressurization tests on various Zircaloy cladding tubes irradiated at different reactor conditions. The cause of the increase in the SCC susceptibility at neutron fluences above 10²⁰ n/cm² would be ascribed to radiation anneal hardening phenomenon and resultant inhomogeneous incipient deformation characterized by dislocation channels.     

Deuterium diffusion in LiOH–water-corroded oxide layer of zirconium alloys

In order to clarify the hydrogen diffusion mechanism in the oxide layer of zirconium alloys, in situ hydrogen isotope diffusion in the oxide layer has been examined. The zirconium alloys used were Zircaloy-2, GNF-Ziron (Zircaloy-2 type alloy with high iron content) and VB (zirconium-based alloy with high iron and chromium contents). They were corroded in 1 or 0.1 M LiOH-containing water at 563 K, producing oxide layers of 1.1–2.1 μm in thickness. The diffusion experiments were carried out in the temperature range from 488 to 633 K by using a combined technique of deuterium plasma exposure and nuclear reaction analysis for D (³He,p)⁴He reaction. From the transient deuterium profiles in the oxide layers, it was concluded the LiOH–water-corroded oxides had a single-layer structure, which was in contrast to the double-layer structure previously observed in steam-corroded oxide layers. The diffusion coefficients in the 1 M LiOH–water-corroded oxides evaluated from the deuterium profiles were smaller in the order of Zircaloy-2 > GNF-Ziron > VB at 573 K. For the 0.1 M LiOH–water-corroded oxide of GNF-Ziron, the diffusivity was lower than that of the 1 M LiOH–water-corroded oxide by a factor of 1/4. The present diffusion coefficients of the 1 M LiOH–water-corroded oxides of GNF-Ziron and VB were approximately 7 times larger than the previous data of the corresponding steam-corroded oxides. The deuterium diffusion properties in the oxides of the three alloys obtained in the in situ experiment were roughly consistent with their hydrogen absorption performances in the LiOH–water-corrosion tests, as well as in the previous steam corrosion tests.     

Application of Zeolites to Remove Iodine from Dissolver Off-Gas, (I)

The oxidation behavior of Zr(Fe,Cr)₂ precipitates being present at the surface of Zircaloy-4 was examined with microprobe Auger electron analysis, focussing attention on the oxidation behavior of chromium and iron in the early stage of oxidation where the oxide film was coherent. Chromium formed a thin oxide layer at the top surface of the zirconium oxide film of precipitate and remained in a metallic state inside the oxide film. Iron was oxidized via dissolution in the matrix zirconium oxide near the top surface and remained in a metallic state inside the oxide film. Such variety of chemical state of chromium and iron with depth in the oxide film was attributed to the existence of oxygen potential gradient in the oxide film.     

A TEM study of the C15 type Zr(CrFe)2 Laves phase in Zircaloy-4

The intermetallic precipitates in Zircaloy-4 have been identified as the C15 type Zr(CrFe)₂ Laves phase using high order Laue zones and series diffraction patterns from the transmission electron microscope. A model for the transformation from the α-Zr matrix to the Zr(CrFe)₂ Laves phase has been constructed using the data on orientation relationships obtained from TEM diffraction patterns. A defect structure of R-type stacking variety has been found in the Zr(CrFe)₂ particles.     

Morphology and composition of second phase particles in zircaloy-2

Intermetallic particles in Zircaloy-2 are analyzed for morphology (shape, size, distribution, etc.) and for crystal structure by transmission electron microscopy. Chemical composition is semi-quantitatively evaluated in a scanning transmission electron microscope. Several types of morphologically distinct particles are identified by shape, size, and substructure. All particles are found to be either nickel-bearing or chromium-bearing. Clusters containing several particles are often observed. The nickel-bearing particles are identified as a tetragonal Zr₂Ni-type phase where iron partially substitutes for some Ni giving an approximate composition of Zr₂Ni_0.4Fe_0.6. The chromium-bearing particles are found to be a hexagonal ZrCr₂-type phase where iron partially substitutes for some Cr giving an approximate composition of ZrCr_1.1Fe_0.9. Essentially all the iron is contained in these two kinds of particles, and no iron-zirconium particles are found.     

Creep behavior of niobium-modified zirconium alloys

Zirconium (Zr) alloys remain as the main cladding materials in most water reactors. Historically, a series of Zircaloys were developed, and two versions, Zircaloy-2 and -4, are still employed in many reactors. The recent trend is to use the Nb-modified zirconium alloys as the Nb addition improves cladding performance in various ways, most significant being superior long term corrosion resistance. Hence, new alloys with Nb additions have recently been developed, such as Zirlo^™2 and M5^™3. Although it is known that creep properties improve, there have been very few data available to precisely evaluate the creep characteristics of new commercial alloys. However, the creep behavior of many Nb-modified zirconium alloys has been studied in several occasions. In this study, we have collected the creep data of these Nb-modified alloys from the open literature as well as our own study over a wide range of stresses and temperatures. The data have been compared with those of conventional Zr and Zircaloys to determine the exact role Nb plays. It has been argued that Nb-modified zirconium alloys would behave as Class-A alloys (stress exponent of 3) with the Nb atoms forming solute atmospheres around dislocations and thus, impeding dislocation glide under suitable conditions. On the other hand, zirconium and Zircaloys behave as Class-M alloys with a stress exponent of ?4, attesting to the dislocation climb-controlled deformation mode.     

High-temperature internal friction in polycrystalline Zirconium and Zircaloy-4

The data on the high-temperature internal friction of zirconium and zirconium alloys are reviewed and new results on zirconium and Zircaloy-4, measured at low and at intermediate frequencies, are presented. It is shown that the damping spectrum of pure zirconium, for annealed polycrystals, shows a peak probably related to relaxation of grain or subgrain boundaries. The data on Zircaloy-4 show two peaks: one near the grain-boundary peak in the pure metal and another one at a higher temperature. Possible mechanisms for these peaks are discussed. Finally, the high-temperature internal friction background of zirconium and zirconium alloys is analyzed and, for Zircaloy-4, the apparent activation enthalpy is found to be related to the grain size.     

Effect of alloying elements on proclivity of zirconium alloys for nodular corrosion

An approach to assessing the effect of alloying elements on the proclivity of zirconium alloys for nodular corrosion is described. The approach is based on an analysis of the stability of the corrosion front. The results of a simplified analysis for iron and nickel additives to Zircaloy-2 and -4 are in agreement with the experimental data. The approach described can also be used for zirconium-niobium alloys, but this requires taking account of the mutual effect of the atoms of niobium and other alloying substances. For detailed quantitative assessments of the effect of alloying elements on the proclivity of zirconium alloys for nodular corrosion, the stability analysis must be expanded and including the use of self-consistent numerical models which take account of radiation effects and oxygen transfer through the oxide film. Translated from Atomnaya énergiya, Vol. 106, No. 2, pp. 94–99, February, 2009.