Zircaloy Corrosion Research Articles

Development of a Zr-Nb-H-O reactive force field for molecular dynamics simulations of in-reactor corrosion

The corrosion of Zircaloy fuel cladding is one of the crucial issues during the operation of pressurized water reactors (PWR). In this work, a Zr-Nb-H-O reactive force field (ReaxFF) is constructed to study the corrosion mechanisms and irradiation effect of Zr-Nb alloys at the atomic landscape. This ReaxFF can describe the interactions between water dissociation products and Zr-Nb alloys, and the stability as well as diffusion properties of irradiation defects in Zr bulk. The molecular dynamics (MD) simulations based on the present ReaxFF show that Nb can thicken the suboxide layer during the corrosion process, as well as the promotion effect of irradiation on corrosion varies under different PKA (Primary Knock-on Atom) energies. The ReaxFF described here has the potential to be a new tool for understanding the in-reactor corrosion mechanism under the irradiation environment.

The role of secondary phase precipitates in zircaloy corrosion: From point defects to local percolation

A thin oxide film on NZ alloy (Zr-1Sn-0.3Nb-0.3Fe-0.1Cr) was formed to investigate the correlation of microstructure and corrosion behavior at the initial stage by directly identifying defect-enriched sites. The sites of high conductivity are revealed and labelled by silver deposition, and in the further investigation of microstructure secondary phase precipitates (SPPs) interconnected with micro cracks are found below. Enrichment of oxygen-18 and deuterium in SPPs after isotopic tracing by secondary corrosion highlights their influence on carrier diffusion. Thus, the percolation for corrosion media transportation is reached locally by SPPs interconnected with micro cracks.

Corrosion of Zircaloy-4, Zr-1Nb and FeCrAl alloys in deaerated water at 330 °C and 14 MPa: Superior corrosion resistance and corrosion mechanism of FeCrAl alloy by experiments and molecular dynamics simulations

After the Fukushima explosion, the research and development of accident tolerant fuel (ATF) cladding has become a hot spot in the nuclear field. In this work, corrosion behavior of Zircaloy-4, Zr-1Nb and FeCrAl alloys was investigated in deaerated water at 330 °C and 14 MPa. All alloys were found to lose weight. In Li/B water, the corrosion resistance followed an order of FeCrAl > Zr-1Nb > Zircaloy-4. Local corrosion occurred on the Zircaloy-4 surface and developed towards the interior of the matrix, leading to selective oxidation. A certain thickness of oxide layer was formed on the surface of Zr-1Nb, but the distribution was uneven. Significant oxide exfoliation made Zr-1Nb display complex defects. The oxide film of FeCrAl alloy had a double-spinel structure. Low surface adsorption energy and weak interaction between Fe and O atoms is responsible for superior corrosion resistance of FeCrAl alloy.

Corrosion of FeCrAl alloys used as fuel cladding in nuclear reactors

The safety of nuclear power plants has put forward new requirements for the corrosion resistance of fuel cladding with the advancement of industrialization. FeCrAl alloys are expected to be a substitute for Zircaloy. This work summarizes the corrosion of Zircaloy and presents a comprehensive review on the corrosion behavior of FeCrAl alloys including oxidation kinetics, corrosion processes and influencing factors in nuclear reactors. Corrosion performance of Zircaloy in nuclear water chemistry is definite while influencing mechanisms of relevant parameters (e.g. coolant pH) are still unclear. A small change of LiOH concentration may cause contrary effect on corrosion behavior. Different from Zircaloy cladding, FeCrAl cladding shows various kinetics in corrosive environments. A single, duplex or triple oxide layer is all possibly observed. Specific corrosion behavior depends on the properties of water chemistry and chemical reactions of main alloying elements. Developing challenges and prospects of FeCrAl alloys are concluded to expand its commercial scale in nuclear fuel cladding. For successful commercial application of FeCrAl cladding, a systematic evaluation considering corrosion resistance and other properties (e.g., mechanical strength, thermal hydraulic characteristics and neutron economy) is required. • Corrosion behavior of Zircaloy in typical nuclear water chemistry is presented. • Kinetics, process and influencing factors of FeCrAl alloys corrosion are reviewed. • Comparison and similarity of corrosion between Zircaloy and FeCrAl are clarified. • Developing challenges and prospects of FeCrAl alloys are proposed.

Influence of light ion irradiation of the oxide layer on the oxidation rate of zircaloy-4: Irradiation of the post-transition oxide layer

The corrosion behavior of Zircaloy-4 fuel cladding in nuclear reactor is affected by irradiation damage. The irradiation effect of the post-transition oxide layer on the Zircaloy-4 corrosion rate has been investigated using protons. As previously observed on pre-transition oxide layer, irradiation defects increase the oxidation rate of the alloy up to around 30 days in autoclave in agreement with the irradiation defect annealing characterized by Raman spectroscopy. The model proposed for pre-transition regime is able to describe the oxidation rate after irradiation of post-transition oxide layers.

14C Exposure from Disposal of Radioactive Waste Compared to14C Exposure from Cosmogenic Origin

ABSTRACTThe potential14C (carbon-14, radiocarbon) flux from disposal of14C containing waste into air is compared with the natural14C emanation rate from soil in order to put the14C hazard potential from disposal of this waste in perspective with the14C exposure from cosmogenic origin. Chemical corrosion of neutron irradiated metals, steel and Zircaloy, is bounded by diffusion of water through a thermodynamically stable metal-oxide layer and dissolution of this metal-oxide in a nuclear plant. Many countries process radioactive waste for disposal using cementitious materials, an acknowledged end-point management technique for this waste. The metal-oxides are also stable when these waste forms are embedded in cementitious materials. The14C release rate from this Zircaloy at these alkaline and reducing conditions is comparable to the natural14C emanation rate from soil into air. Neutron irradiated graphite and spent ion exchange resins are chemically inert and therefore other release mechanisms need to be assumed. Radiolytic corrosion is used to determine the14C release rate from this graphite. Moreover, ion exchange—with ingressing anionic species that have a higher affinity than contained anionic14C—is proposed as a release mechanism for these resins.

Improve Corrosion Resistance of Zircaloy by Copper–Graphene Nanocomposite Coatings

In this study, the corrosion behavior of Zircaloy was investigated in the presence and absence of copper–graphene nanocomposites coating. The coating was prepared employing Hummers’ and electrochemical reduction methods. The morphology of copper–graphene nanocomposites coating was studied using scanning electron microscopy (SEM). Corrosion behavior was investigated employing dynamic polarization and electrochemical impedance spectroscopy (EIS) tests in a solution containing lithium hydroxide (LiOH), boric acid (H3BO3), and deionized water. The results showed that corrosion resistance of Zircaloy increased with introduction of copper–graphene nanocomposites coatings. The lowest corrosion rate was attained in the Zircaloy with copper–graphene nanocomposites coating (corrosion rate: 0.040 mm/year). An approximately 20 times decrease in the corrosion rate was observed in the Zircaloy with copper–graphene nanocomposites coating when compared to the un-coated Zircaloy (corrosion rate: 0.831 mm/year).

Phase field modeling of irradiation-enhanced corrosion of Zircaloy-4 in PWRs

A phase field model (HOGNOSE) has been developed to address the incomplete understanding of early stage irradiation-enhanced corrosion of Zircaloy-4 in pressurized water reactors (PWRs). HOGNOSE uses an effective charge density to represent space charge in the oxide, which is effectively removed under irradiation by iron doping after secondary phase particle amorphization. Accounting for the temperature and neutron flux dependence on amorphization and showing the large impact of doping on both electron and oxygen transport, HOGNOSE models irradiation-enhanced corrosion up to ten microns over 270–330 °C and prototypical PWR neutron fluxes.

Release and speciation of carbon from Zircaloy-4 in anaerobic and highly alkaline conditions: Comparison of simple immersion and potentiostatic corrosion tests

ABSTRACTThe gas release and speciation of carbon species from irradiated and unirradiated Zircaloy-4 samples, representative for the fuel cladding as used in Belgian nuclear power plants, were studied in a saturated Ca(OH)2 solution in anaerobic conditions. This environment is relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level nuclear waste. To achieve this, we performed simple immersion and potentiostatic corrosion tests. Potentiodynamic polarization curves, recorded prior to the potentiostatic tests, revealed that irradiation seems to induce changes on the Zircaloy-4 corrosion behavior, such as a shift of the corrosion potential. Potentiostatic corrosion tests on unirradiated Zircaloy-4 provided a corrosion rate of ~54 nm/yr over a 7 day-experiment, whilst a corrosion rate of only ~4 nm/yr was calculated for the irradiated sample. Gas chromatography revealed that during simple immersion tests, which lasted 195 days, hydrogen, methane, ethane, and CO2 were produced, with methane being the major compound. Assuming that all carbon released from the metal was transformed into gaseous compounds, this yields to a corrosion rate ranging from 57 to 84 nm/yr for the irradiated sample. However, caution has to be taken on these corrosion rate and more tests should be performed to confirm these results.

A novel mechanism for nodular corrosion of Zircaloy-4 corroded in 773 K superheated steam

Corrosion resistance of Zircaloy-4 plate specimens in 773K/10.3MPa superheated steam is anisotropic. The rolling surface (SN) mainly suffers uniform corrosion, while the surface perpendicular to transversal direction (ST) and the surface perpendicular to rolling direction (SR) mainly suffer nodular corrosion. A novel mechanism of nodular corrosion related to Sn segregation at the interface between oxide and matrix (O/M) is proposed. The anisotropic nodular corrosion results from the anisotropy of Sn segregation at the O/M interfaces formed on SN, SR and ST surfaces with different grain orientations. Finally, some reported results about nodular corrosion in zirconium alloys are discussed.

Effect of hydrogen uptake on the electrochemical corrosion of N18 zircaloy under gamma irradiation

It has been well recognized that dramatic hydrogen uptake occurred in zircaloy after kinetic transition and porous structure was observed subsequently due to phase transformation of tetragonal to monoclinic zirconia. Therefore, how hydrogen solute and gamma-induced capillary-embedded hydrolysis influence the corrosion of zircaloy is an intriguing issue. In this work, the effect of hydrogen uptake and gamma irradiation on corrosion of N18 zircaloy was studied. Raman spectra and atomic force microscopy (AFM) were employed to analyse phase structure and surface morphology. Potentiodynamic polarization and electrochemical impedance spectroscopy were utilized to qualitatively evaluate the electron transfer properties of the oxide film formed on the zircaloy surface after corrosion. The depth profile and surface chemical states of involving elements were analysed by auger electron spectroscopy (AES) and X-ray photoelectron spectroscopy (XPS), respectively. It was found that hydrogen permeation can decline the integrity and impedance semicircle of the oxide films, the more the hydrogen uptake is; the smaller magnitude of impedance will be. In view of the gamma irradiation, it is demonstrated that it promotes the corrosion rate slightly. Based on the irradiation theory and existing phenomena, the underlying mechanism is proposed.

Carbon 14 Distribution in Irradiated BWR Fuel Cladding and Released Carbon 14 after Aqueous Immersion of 6.5 Years

Spent fuel cladding which is highly activated and strongly contaminated is expected to be disposed of in an underground repository. A typical activation product in the activated metal waste is carbon 14 (14C), which is mainly generated by the 14N(n,p)14C reaction and produces a significant exposure dose due to the large inventory, long half-life (5730 years), rapid release rate, and the speciation and consequent migration parameters. In the preliminary Japanese safety case, the release of radionuclides from the metal matrix is regarded as the corrosion-related congruent release, and the cladding oxide layer is regarded as a source of instant release fraction (IRF).In the present work, specific activity of 14C was measured using an irradiated BWR fuel cladding (Zircaloy-2, average rod burnup of 41.6 GWd/tU) which has an external oxide film having a thickness of 25.3 μm. The 14C specific activity of the base metal was 1.49×104 Bq/g, which in the corresponding burnup is comparable to values in the existing literature, which were obtained from various irradiated claddings. Although the specific activity in oxide was 2.8 times the base metal activity due to the additive generation by the 17O(n,α)14C reaction, the 14C abundance in oxide was less than 10% of total inventory. A static leaching test using the cladding tube was carried out in an air-tight vessel filled with a deoxygenated dilute NaOH solution (pH of 12.5) at room temperature. After 6.5 years, 14C was found in each leachate fraction of gas phase and dissolved organics and inorganics, the total of which was less than 0.01% of the 14C inventory of the immersed cladding tube. A simple calculation based on the congruent release with Zircaloy corrosion has suggested that the 96.7% of released 14C was from the external oxide layer and 3.3% was from the base Zircaloy metal. However, both the 14C abundance and the low leaching rate suggests that 14C in oxide does not have a significant impact on the IRF in the safety case.

Effects of seawater components on radiolysis of water at elevated temperature and subsequent integrity of fuel materials

Effects of seawater components on radiolysis of water at elevated temperature have been studied with a radiolysis model and a corrosion test under gamma-ray irradiation conditions to evaluate the subsequent influence on integrity of fuel materials used in an advanced boiling water reactor. In 2011, seawater flowed into the nuclear power plant system of the Hamaoka Nuclear Power Station Reactor No. 5 during the plant shutdown operation. The reactor water temperature was 250 °C and its maximum Cl− concentration was ca. 450 ppm when seawater was mixed with reactor water. The radiolysis model predicted that the main radiolytic species were hydrogen, oxygen and hydrogen peroxide. Concentrations of radiolytic products originating from Cl− and other seawater components were found to be rather low. The dominant product among them was ClO3− and its concentration was found to be below 0.01 ppm for a 105 s irradiation period. No significant corrosion of zircaloy-2 and 316L stainless steel was found in the corrosion test. These results led to the conclusion that the harmful influence of radiolytic products originating from seawater components on integrity of fuel materials must be smaller than that of Cl− which is the main ionic species in seawater.

Mechanism of corrosion of zirconium hydride and impact of precipitated hydrides on the Zircaloy-4 corrosion behaviour

In Pressurized Water Reactors, zirconium hydrides precipitate in the matrix and could increase the oxidation rate of the claddings. To understand their effect, corrosion tests, TEM and μ-XRD analyses have been performed. This work showed that the oxidation rate and the oxygen diffusion coefficient in the oxide formed on massive hydride are much greater than those of Zircaloy-4. Moreover, oxide characterizations indicated an additional phase indexed as the sub-oxide Zr3O between the oxide film and the massive hydride. Finally, the hydrogen of the hydrides is not incorporated in the oxide during the corrosion process.

Influence of light ion irradiation of the oxide layer on the oxidation rate of Zircaloy-4

A strong increase of the oxidation rate of Zircaloy-4 fuel cladding is observed for burnups above 35GWd/MtU in PWR. This kinetic acceleration could be in large part due to the irradiation damage. To quantify the effect of the oxide layer irradiation on the Zircaloy-4 corrosion rate, a new approach was used. Defects created by light ion irradiation increase the oxidation rate of the alloy up to 23 days in autoclave in agreement with the irradiation defect annealing characterized by RAMAN spectroscopy. A simple model taking into account the annealing of the irradiation defect is proposed to explain the results.

Corrosion Tests of Zircaloy Hull Waste to Confirm Applicability of Corrosion Model and to Evaluate Influence Factors on Corrosion Rate under Geological Disposal Conditions

ABSTRACTCorrosion behavior is a key issue in the assessment of disposal performance for activated waste such as spent fuel assemblies (i.e., hulls and end-pieces) because corrosion is expected to initiate radionuclide (e.g., C-14) leaching from such waste. Because the anticipated corrosion rate is extremely low, understanding and modeling Zircaloy (Zry) corrosion behavior under geological disposal conditions is important in predicting very long-term corrosion. Corrosion models applicable in the higher temperature ranges of nuclear reactors have been proposed based on considerable testing in the 523−633 K temperature range.In this study, corrosion tests were carried out to confirm the applicability of such existing models to the low temperature range of geological disposal, and to examine the influence of material, environmental, and other factors on corrosion rates under geological disposal conditions. A characterization analysis of the generated oxide film was also performed.To confirm applicability, the corrosion rate of Zry-4 in pure water with a temperature change from 303 K to 433 K was obtained using a hydrogen measuring technique, giving a corrosion rate for 180 days of 8 × 10-3 μm/y at 303 K.To investigate the influence of various factors, corrosion tests were carried out. The corrosion rates for Zry-2 and Zry-4 were almost same, and increased with a temperature increase from 303 K to 353 K. The influence of pH (12.5) compared with pure water was about 1.4 at 180 days at 303 K.

Long-term Corrosion of Zircaloy-4 and Zircaloy-2 by Continuous Hydrogen Measurement under Repository Condition

ABSTRACTCorrosion behavior is a key issue for the waste disposal of irradiated metals, such as hulls and endpieces, and is considered to be a leaching source of radionuclides including C-14. However, little information about Zircaloy corrosion in anticorrosive conditions has been provided.In the present study, long-term corrosion tests of Zircaloy-4 and Zircaloy-2 were performed in assumed disposal conditions (dilute NaOH solution, pH 12.5, 303 K) by using the gas flow system for 1500 days. The corrosion rate, which was determined by measuring gaseous hydrogen and the hydrogen absorbed in Zircaloy, decreased with immersion time and was lower than the value of 2×10−2 μm/y used in performance assessment (1500-day values: 5.84×10−3 and 5.66×10−3 μm/y for Zircaloy-4, 1000-day values: 8.81×10−3 μm/y for Zircaloy-2). The difference in corrosion behavior between Zircaloy 4 and Zircaloy-2 was negligible. The average values of the hydrogen absorption ratios for Zircaloy-4 and Zircaloy-2 during corrosion were 91% and 94%, respectively.The hydrogen generation kinetics of both gas evolution and absorption into metal can be shown by a parabolic curve. This result indicates that the diffusion process controls the Zircaloy corrosion in the early corrosion stage of the present study, and that the thickness of the oxide film in this stage is limited to approximately 25 nm and may therefore be in the form of dense tetragonal zirconia.

Specific Aspects of Internal Corrosion of Nuclear Clad Made of Zircaloy

In PWR, the Zircaloy based clad is the first safety barrier of the fuel rod, it must prevent the dispersion of the radioactive elements, which are formed by fission inside the UO2 pellets filling the clad. We focus here on internal corrosion that occurs when the clad is in tight contact with the UO2 pellet. In this situation, with temperature of 400 °C on the internal surface of the clad, a layer of oxidised Zircaloy is formed with a thickness ranging from 5 to 15 µm. In this paper, we will underline the specific behaviour of this internal corrosion layer compared to wet corrosion of Zircaloy. Simulations will underline the differences of stress field and their influences on corresponding dissolved oxygen profiles. The reasons for these differences will be discussed as function of the mechanical state at inner surface of the clad which is highly compressed. Differences between mechanical conditions generated by an inner or outer corrosion of the clad are studied and their influences on the diffusion phenomena are highlighted.

Understanding Crack Formation at the Metal/Oxide Interface During Corrosion of Zircaloy-4 Using a Simple Mechanical Model

Abstract It has been established in previous works that corrosion kinetics in primary water of various zirconium alloys are periodic. Each period is associated with a layer of cracks parallel to the metal-oxide interface. These observations have been made either in autoclave or in pile. This indicates that corrosion processes in autoclave and under irradiation are of similar nature though their absolute kinetics might be different. Taking advantage of this correlation between cracks and corrosion kinetics, the present work aims at identifying the main microstructural parameters controlling cracks appearance in the oxide layer under well-controlled conditions. In order to achieve this, Zircaloy-4 was heat-treated to obtain various metallurgical states (stress-relieved versus recrystallised with different grain sizes) followed by corrosion tests in primary water. The key metallurgical parameters for the various conditions have been analysed (texture, precipitate sizes and grain sizes and distributions) using electron microscopy and synchrotron X-ray diffraction techniques. Corrosion kinetics of the various Zircaloy-4 microstructures are distinct as expected from the literature. Crack morphology in the oxide layer has been analysed and quantified using a dual beam scanning electron microscope/focused ion beam. Crack layers are evident even at small scale of observation. Three dimensional (3D) images of the oxide structure are presented. Cracks observed in this way are typically penny-shaped with a radius of about 100 nm. Near the metal-oxide interface, they are mainly found at the top of metal protrusions in the oxide. The roughness of the metal-oxide interface was measured. It does not exhibit any periodicity. The residual stresses in the oxide layers were measured by high energy (44 keV) synchrotron X-ray diffraction in transmission mode. Large compressive stresses (∼−1 GPa), changing with the metallurgical state and through the oxide scale thickness, were measured. The residual stresses in the oxide layers were measured by high energy (44 keV) synchrotron X-ray diffraction in transmission mode. Large compressive stresses (∼−1 GPa), changing with the metallurgical state and through the oxide scale thickness, were measured. A model of the oxide breaking at the point of transition has been developed. It is based on mechanical considerations and the existence of compressive stress in the oxide layer.

Understanding of Hybriding Mechanisms of Zircaloy-4 Alloy during Corrosion in PWR Simulated Conditions and Influence of Zirconium Hybrides on Zircaloy-4 Corrosion

Les alliages de Zirconium sont couramment utilises comme materiaux de gainage dans les Reacteurs a Eau Pressurisee. Immerges dans le milieu primaire, alors que la face externe des crayons s'oxyde, une partie de l'hydrogene impliquee dans la reaction d'oxydation penetre dans le materiau et vient precipiter dans la matrice metallique sous la forme d'hydrures fragilisants, une fois la limite de solubilite de l'hydrogene atteinte a la temperature d'irradiation. Ce phenomene d'hydruration est l'un des facteurs limitant l'utilisation des alliages de type Zircaloy-4 (Zr-Fe-Cr-Sn-O). En tant que premiere barriere de confinement des produits de fission, l'integrite mecanique des crayons de combustible est essentielle, et la minimisation de la sensibilite de cet alliage a la prise d'hydrogene est requise. Cette demarche necessite notamment la comprehension des mecanismes regissant la prise d'hydrogene du materiau, ainsi que l'influence des hydrures sur son comportement en corrosion. Concernant l'etude des mecanismes de prise d'hydrogene, des echanges isotopiques en eau lourde primaire a 360°C ont ete entrepris et ont mene a l'identification, dans le film d'oxyde, a l'etape limitant la diffusion de l'espece hydrogenee. Afin d'y estimer le coefficient de diffusion apparent de l'hydrogene, nous nous sommes tout d'abord appuyes sur des profils SIMS de penetration du deuterium dans la partie dense du film d'oxyde. Dans un second temps, des estimations ERDA de la teneur en hydrogene presente dans l'oxyde associees a des dosages par fusion nous ont permis de determiner le flux d'hydrogene absorbe par l'alliage puis de deduire un coefficient de diffusion apparent a 360°C. Finalement, ces deux methodes convergent vers des valeurs analogues (entre 1.10 -14 cm 2 /s et 6.10 -14 cm 2 /s), en accord avec les donnees de la litterature. Concernant l'impact des hydrures sur la corrosion du Zircaloy-4, plusieurs echantillons pre-hydrures et vierges de reference ont ete oxydes simultanement en eau primaire a 360°C. La caracterisation des echantillons pre-hydrures a revele certaines differences notables avec les echantillons temoins, comme la presence du sous-oxyde Zr 3 O a l'interface metal/oxyde, une plus faible proportion de zircone quadratique dans l'oxyde, ainsi qu'une diffusion plus rapide de l'oxygene via les courts-circuits de la zircone (MET, μ-DRX, echanges isotopiques en 18 O). De plus, au cours de l'oxydation, l'hydrogene initialement present dans la phase hydruree reste dans la matrice metallique et conduit progressivement a la transformation allotropique δ-ZrH 1,66 → e-ZrH 2 .