Hydrogen Pickup Fraction Research Articles

Oxidation behavior of Zr-2.5Nb alloy exposed to steam in the temperature range of 600–1200 °C

Zr-2.5Nb pressure tube material was subjected to isothermal steam exposures in the range 600–1200 °C for different durations to predict its oxidation behavior during hypothetical accident scenarios. The oxidation exponent obtained from power law fitting of the thermogravimetry (TG) curves deviated from parabolic regime. Nevertheless, oxidation kinetics was described by determining the parabolic rate constants (Kp) in two temperature ranges: Kp = 0.029 exp(-111075/RT) (kg/m2)2/sec for 600–900 °C whereas Kp = 77652.6 exp(-254824/RT) (kg/m2)2/sec for 900–1200 °C. In the tested durations, only the 1000 °C sample has shown breakaway transition, after approximately 105 min and has absorbed the highest amount of hydrogen amongst all samples. The trend in hydrogen pickup fraction (HPUF) was verified by two complementary hydrogen determination methods. Clear differences were identified in the α-Zr(O) layer formed on the samples below and above 900 °C. In the α + β range of 600–900 °C, α-Zr(O) consisted of tiny β precipitates. At and above 900 °C in the fully β range, columnar α-Zr(O) grains were observed.

The effect of varying nickel concentration on the hydrogen pickup fraction and corrosion resistance in Zircaloy-4

Zircaloy-4 samples with various levels of Ni content (0–483 wt. ppm) were exposed in autoclave for up to 2800 days. Samples were destructively and non-destructively measured to determine the weight gain from oxidation and hydrogen contents to better understand the role hydrogen has in increasing oxide growth rates and to study the effect of Ni on the mechanism of hydrogen pickup in Zr alloys. Overall, additions of Ni lead to significantly higher hydrogen pickup of the alloy. As observed previously, at higher weight gains, a small decrease in hydrogen pickup fraction is observed for all alloys tested (Zircaloy-4 with and without Ni). A small beneficial effect of Ni was observed on corrosion rate when controlling for the amount of hydrogen in the material, which potentially indicates that if hydrogen content were controlled via other means (e.g., volume or other alloy additions) Ni may improve corrosion resistance of the material. Results generally support published density functional theory (DFT) hypotheses that point to an increase in hydrogen solubility or catalysis of surface reactions as the mechanism by which Ni increases hydrogen pickup. Data or future corrosion testing of alloys with Ta, V, or Ti additions could further support or refute these theories by testing aspects of the proposed mechanisms.

Understanding the role of Fe, Cr and Ni in Zircaloy-2 with special focus on the role of Ni on hydrogen pickup

Ni as an alloying addition in Zircaloy leads to an increase in hydrogen pick-up fraction. Atomic scale simulations of tetragonal ZrO2, based on density functional theory, are used to identify a possible mechanism for this observation. First, defect formation energies associated with Ni but also Fe and Cr are used to predict relative defect cluster and defect charge concentrations using Brouwer diagrams. At low oxygen partial pressures (PO2), expected in the vicinity of the oxide metal interface, a cluster consisting of an oxygen vacancy adjacent to a charge neutral Ni0 atom is identified as the most populous cluster. Further simulations show that a hydrogen molecule will dissociate in the vicinity of this cluster. No other cluster is both sufficiently populous and acts in this way. This differentiates Ni from the other alloying elements.

The effect of Nb on the corrosion and hydrogen pick-up of Zr alloys

Zr-Nb alloys are known to perform better in corrosion and hydrogen pick-up than other Zr alloys but the mechanism by which this happens is not well understood. Atomistic simulations using density functional theory of both tetragonal and monoclinic ZrO2 were performed, with intrinsic defects and Nb dopants. The overall defect populations with respect to oxygen partial pressure were calculated and presented in the form of Brouwer diagrams. Nb is found to favour 5 + in monoclinic ZrO2 at all partial pressures, but can exist in oxidation states ranging from 5 + to 3 + in the tetragonal phase. Nb5+ is charge balanced by Zr vacancies in both phases, suggesting that contrary to previous assumptions, Nb does not act as an n-type dopant in the oxide layer. Clusters containing oxygen vacancies were considered, Nb2+ was shown to exist in the tetragonal phase with a binding energy of 2.4 eV. This supports the proposed mechanism whereby low oxidation state Nb ions (2 + or 3+) charge balance the build-up of positive space-charge in the oxide layer, increasing oxygen vacancy and electron mobility, leading to near-parabolic corrosion kinetics and a reduced hydrogen pick-up. Previous experimental work has shown that tetragonal ZrO2 transforms to the monoclinic phase during transition, and that during transition a sharp drop in the instantaneous hydrogen pick-up fraction occurs. The oxidation of lower charge state Nb defects to Nb5+ during this phase change, and the consequent temporary n-doping of the oxide layer, is proposed as an explanation for the drop in hydrogen pick-up during transition.

In-situ electrochemical impedance spectroscopy measurements of zirconium alloy oxide conductivity: Relationship to hydrogen pickup

Hydrogen pickup during nuclear fuel cladding corrosion is a critical life-limiting degradation mechanism for nuclear fuel. Following a program dedicated to zirconium alloys, corrosion, it has been hypothesized that oxide electronic resistivity determines hydrogen pickup. In-situ electrochemical impedance spectroscopy experiments were performed on Zircaloy-4 and Zr-2.5Nb alloys in 360°C water. The oxide resistivity was measured as function of time. The results show that as the oxide resistivity increases so does the hydrogen pickup fraction. The resistivity of the oxide layer formed on Zircaloy-4 is higher than on Zr-2.5Nb, resulting in a higher hydrogen pickup fraction of Zircaloy-4, compared to Zr-2.5Nb.

Hydrogen pickup measurements in zirconium alloys: Relation to oxidation kinetics

The optimization of zirconium-based alloys used for nuclear fuel cladding aims to reduce hydrogen pickup during operation, and the associated cladding degradation. The present study focuses on precisely and accurately measuring hydrogen pickup fraction for a set of alloys to specifically investigate the effects of alloying elements, microstructure and corrosion kinetics on hydrogen uptake. To measure hydrogen concentrations in zirconium alloys two techniques have been used: a destructive technique, Vacuum Hot Extraction, and a non-destructive one, Cold Neutron Prompt Gamma Activation Analysis. The results of both techniques show that hydrogen pickup fraction varies significantly with exposure time and between alloys. A possible interpretation of the results is that hydrogen pickup results from the need to balance charge. That is, the pickup of hydrogen shows an inverse relationship to oxidation kinetics, indicating that, if transport of charged species is rate limiting, oxide transport properties such as oxide electronic conductivity play a key role in the hydrogen pickup mechanism. Alloying elements (either in solid solution or in precipitates) would therefore impact the hydrogen pickup fraction by affecting charge transport.

Impact of additives on zirconium oxidation by water: mechanistic insights from first principles

Zirconium alloys are widely used as cladding materials employed to contain the fission fuel in nuclear power plants. A limiting factor for fuel longevity is the corrosion property of the zirconium alloys. In the reactor, the main corrodent is water. The oxide forms thermodynamically during hydrogen evolution. Due to the corrosion mechanism, a fraction of the hydrogen is transferred to the alloy. It has long been known that the alloying elements actually control the hydrogen pick-up fraction, HPUF. A mechanism that explains these observations by means of density functional theory calculations is presented and validated. A hydroxylated grain boundary model decorated by various transition metal, TM, ions is employed to study the dependence of the hydrogen evolution reaction, HER, on the choice of TM ion and spin state along the hydride–proton recombination pathway. The efficiency of the system to utilize the overpotential for hydrogen evolution, originating from the overall corrosivity of the alloy, is found to be decisive for the HPUF. A dual origin of the detrimental effects of Co and Ni additives on the HPUF is identified.

Corrosion of M5 in PWRs: Quantification of Li, B, H and Nb in the Oxide Layers Formed Under Different Conditions

Abstract Until now, most of the detailed characterizations of the M5 corrosion behaviour were performed under standard PWR operating conditions, under moderate Li content and moderate temperature of the primary coolant. In this study, in addition to these standard conditions, two demanding operating conditions were explored: increased Li chemistry and elevated temperature. The objective is to establish whether these more demanding conditions have an impact on the structure of the oxide layers formed, on Nb, Li and B contents in these layers and on Hydrogen pickup of the cladding. The structure of oxide layers was studied by microscopy, the Nb content and distribution by Electron Probe Micro Analysis, the Li and B contents and distributions by Nuclear Reaction Analysis and the hydrogen pickup by gas extraction. It was observed that the stability of the corrosion behaviour of M5 is not affected by increased Li or elevated temperature conditions. The hydrogen pickup fraction of M5 is not modified by increased Li conditions or by irradiation temperature with measured contents (<100 ppm) below the solubility limit at irradiation temperature (586–631 K range). Moreover, no significant release of Nb from the oxide in the primary coolant was observed, whatever the operating conditions can be, with quantification accuracy below 0.05%. The respective Li and B contents reached in oxides are of the same order of magnitude under the all three conditions, with average Li content around 15 ppm and average B content varying from 70 to 140 ppm, corresponding to a range where enhanced corrosion is neither observed, nor expected.

Optimization of Zry-2 for High Burnups

Abstract Two major high burnup boiling water reactor (BWR) issues today are (1) shadow corrosion (hydrogen) induced fuel channel bowing and (2) accelerated hydrogen pickup fraction, HPUF of fuel claddings. Both these issues occur only at high burnups. Several factors impact the tendency for fuel channel bowing, the most recent one that is responsible for the current issue of difficulties to insert control rods is due to differences in hydrogen pickup of the different facing surfaces of the fuel channel. Accelerated HPUF of fuel cladding at high burnups may result in reducing fuel clad integrity during reactivity initiated accidents and cask drop accident during dry storage conditions. Some fuel vendors are now shifting from Zry-2 to Zry-4 in structural components to mitigate the fuel channel bowing issue. All BWR fuel vendors are also looking into new alloys (other than Zry-2 and Zry-4) for structural materials. However, one should keep in mind that it is imperative that a mechanistic understanding of the accelerated HPUF is developed to understand the accelerated HPUF tendency in Zry-2. Without such an understanding, it is not possible to rule out that accelerated HPUF can occur also for the new alloys under certain in-pile conditions. This paper summarizes the state of knowledge on the accelerated HPUF and provides suggestions on (1) the mechanism responsible for the accelerated HPUF of Zry-2 at high burnups and (2) how the currently used Zry-2 material for fuel rods and structural materials in BWRs can be modified to delay the accelerated HPUF at high burnups.

In PWR Comprehensive Study of High Burn-Up Corrosion and Growth Behavior of M5® and Recrystallized Low-Tin Zircaloy-4

Abstract This paper is dealing with the corrosion and growth behavior of M5® and recrystallized low tin Zircaloy-4 irradiated as stress-free tubes in conditions representative of grids and guide-tubes of PWR fuel assemblies. The low-tin Zircaloy-4 tubes have reached equivalent burn-ups up to 93 GWd/tU (corresponding to a fluence of 21.0×1025 n⋅m−2, E>1 MeV), and the M5® tubes have reached equivalent burn-ups up to 79 GWd/tU (corresponding to a fluence of 17.1×1025 n⋅m−2). Postirradiation growth was measured by comparing the distances between holes regularly distributed along the rod to the initial measurements. Postirradiation outer diameter oxide thicknesses were measured by eddy currents. Detailed characterizations of oxide layers have been realized through optical microscopy and scanning electron microscopy. Hydrogen uptake has been obtained through global hydrogen content measurements and metallographic examination of hydrides. Finally, the irradiation-induced microstructure of the metallic matrix has been observed by transmission electron microscopy. On these tubes, the free growth reached 1.9 % at 93 GWd/tU on Zircaloy-4 and 0.3 % at 79 GWd/tU on M5®. The external oxide thicknesses are far greater on Zircaloy-4 (∼60 μm at 93 GWd/tU) than on M5® (∼7.5 μm at 79 GWd/tU), and the oxidation rate is eight times higher on Zircaloy-4 than on M5®. Very high hydrogen content is achieved on Zircaloy-4, up to ∼1600 ppm (due to the low wall thickness and two-sided corrosion), whereas the maximum value on M5® is ∼100 ppm (despite the same geometry and corrosion conditions). On both alloys, the hydrogen pickup fraction is of the same order for these experimental empty rods as for previously analyzed fuel rod claddings, which may indicate the absence of a heat flux effect on the hydrogen uptake. Finally, potential correlation between growth, corrosion and hydrogen uptake will be discussed. Taking account of the results obtained on both experimental tubes and fuel rods, the effects of the presence or absence of heat flux through the wall thickness and of hydride rim at the metal/oxide interface are discussed, especially concerning the high burn-up corrosion acceleration on Zircaloy-4.

Irradiated Behavior at High Burnup for HiFi Alloy

Irradiation tests of a BWR advanced Zr alloy (HiFi alloy) and Zircaloy-2 (Zry-2) were carried out in a Japanese commercial reactor and the irradiation performances of the materials were investigated. HiFi alloy and Zry-2 showed excellent resistance to corrosion up to 70 GWd/t, and furthermore, HiFi kept lower hydrogen pickup compared with Zry-2. TEM observation showed that the Fe/(Fe+Cr) ratio of Zr(Fe,Cr)2 type second phase particles (SPPs) for HiFi alloy and Zry-2 tended to decrease as fast neutron fluence increased and to saturate at high fluence. Zr-Fe-Cr SPPs did not completely disappear even for 6 cycles for the irradiated HiFi alloy and Zry-2. In order to clarify the mechanism of hydrogen absorption, an electrochemical technique was used for the oxide film of both materials as part of the out-of-pile test. The relation between the oxide surface potential and the hydrogen pickup fraction was estimated suggesting that the potential difference over the oxide film suppressed hydrogen (proton) diffusion in the oxide film.

Effect of Alloying Elements and Impurities on in-BWR Corrosion of Zirconium Alloys

Abstract 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.

Irradiated Behavior at High Burnup for HiFi Alloy

Irradiation tests of a BWR advanced Zr alloy (HiFi alloy) and Zircaloy-2 (Zry-2) were carried out in a Japanese commercial reactor and the irradiation performances of the materials were investigated. HiFi alloy and Zry-2 showed excellent resistance to corrosion up to 70 GWd/t, and furthermore, HiFi kept lower hydrogen pickup compared with Zry-2. TEM observation showed that the Fe/(Fe+Cr) ratio of Zr(Fe,Cr)2 type second phase particles (SPPs) for HiFi alloy and Zry-2 tended to decrease as fast neutron fluence increased and to saturate at high fluence. Zr-Fe-Cr SPPs did not completely disappear even for 6 cycles for the irradiated HiFi alloy and Zry-2. In order to clarify the mechanism of hydrogen absorption, an electrochemical technique was used for the oxide film of both materials as part of the out-of-pile test. The relation between the oxide surface potential and the hydrogen pickup fraction was estimated suggesting that the potential difference over the oxide film suppressed hydrogen (proton) diffusion in the oxide film.

Developments in PWR cladding corrosion: materials and models

In pressurized water reactors Zircaloy-4 is a standard fuel cladding material. The aim of this paper is to present and evaluate corrosion data generated both in-reactor, and out-of-reactor on PWR claddings made of both Zircaloy-2 and Zircaloy-4 materials. The oxide thickness measurements of cladding tubes irradiated in the Ringhals 3 reactor, and oxide weight gain measurements carried out in Sandvik autoclaves at 400°C, 10.3 MPa clearly show that the stress relief annealed Zircaloy-2 is more corrosion resistant than Zircaloy-4 produced with an identical fabrication route. Furthermore, autoclave tests indicate that the hydrogen pickup fraction of the two alloys is very similar. The obtained data have been evaluated in regard to chemical composition and heat treatment. In addition, computer models, which simulate thermal and hydraulic reactor conditions and corrosion kinetic processes simultaneously, have been used to predict the in-reactor corrosion behaviour of the claddings.

ＢＷＲ用燃料被覆管の耐食性に及ぼすジルカロイ合金組成の影響

Effects of Fe and Ni contents on nodular corrosion susceptibility and hydrogen pick-up of Zircaloy were investigated. Total number of 31 Zr alloys having different chemical compositions; five Zr-Sn-Fe-Cr alloys, eight Zr-Sn-Fe-Ni alloys and eighteen Zr-Sn-Fe-Ni-Cr alloys, were melted and processed to thin plates for the corrosion tests in the environments of a high temperature (510°C) steam and a high temperature (288°C) water.In addition, four 450kg ingots of Zr-Sn-Fe-Ni-Cr alloys were industrially melted and BWR fuel cladding tubes were manufactured through a current material processing sequence to study their producibility, tensile properties and corrosion resistance.Nodular corrosion susceptibility decreased with increasing Fe and Ni contents of Zircaloys. It was seen that the improved Zircaloys having Fe and Ni contents in the range of 0.30[Ni]+0.15[Fe]≥0.045 (w/0) showed no susceptibility to nodular corrosion.An increase of Fe content resulted in a decrease of hydrogen pick-up fraction in both steam and water environments. An increase of Fe and Ni content of Zircaloys in the range of Fe≤0.25w/0 and Ni≤0.1w/0 did not cause the changes in tensile properties and fabricabilities of fuel cladding tube. The fuel cladding tube of improved Zircaloy, containing more amount of Fe and Ni than the upper limit of Zircaloy-2 specification showed no susceptibility to nodular corrosion even in the 530°C steam test.

Hafnium corrosion behavior in high-temperature steam

Corrosion properties for three kinds of hafnium have been examined in 10.5 MPa steam at 773 K. Nuclear grade hafnium formed a shiny black film with a few white nodules, which were found to be monoclinic HfO 2. Hydrogen pick-up fraction during the corrosion amounted to ca. 40% to produce hydrogen dissolved in the hafnium matrix with evolving the remaining 60% hydrogen as H 2 gas. Sponge and crystal bar hafnium were also examined and they showed superior and inferior corrosion resistance, respectively, to nuclear grade hafnium. The corrosion resistance for hafnium increased with increasing the iron content involved in hafnium for these three kinds of hafnium and, correspondingly related to the density of iron-containing second phase particles, which were characterized to be face centered cubic Hf 2Fe. The corrosion mechanism, which was previously proposed for Zircaloy nodular corrosion, was adopted with making minor alterations, to explain the hafnium corrosion properties.