Delayed Hydride Crack Velocity of Zirconium Alloys with the Direction of an Approach to Temperature

The delayed hydride cracking (DHC) tests were conducted on Zr-2.5Nb compact tension specimens with the test temperatures reached by a heating and a cooling. The Zr-2.5Nb specimens were either furnace-cooled or water-quenched after a hydrogen charging treatment to contain 10 to 100 ppm H. On an approach to the test temperatures by a cooling, both the Zr-2.5Nb specimens showed the DHC velocity increasing with an increasing temperature over a temperature range of 100-300°C, irrespective of the cooling rate. However, on an approach to the test temperatures by heating, the furnace-cooled Zr-2.5Nb showed a DHC arrest at temperatures over 180°C and no DHC at 250°C, and the water-quenched ones did have a DHC growth, even at 250°C. Using Kim's DHC model we elucidate the DHC arrest in the furnace-cooled Zr-2.5Nb at temperatures over 180°C and the DHC growth in the water-quenched specimen, even at 250°C, upon an approach by a heating.     

Comment on ‘The first step for delayed hydride cracking in zirconium alloys’ by G.A. McRae et al., J. Nucl. Mater. 396 (2010) 130-143

The aim of this paper is a reply to McRae et al.’s paper entitled “The first step for delayed hydride cracking (DHC) in zirconium alloys” claiming that the first step of DHC is hydrogen diffusion, not nucleation of hydrides as demonstrated by Kim’s new model. Despite the authors’ claim that the crack tip concentration is higher than the bulk concentration due to the stress gradient, their claim violates the thermodynamic principle that the stressed region should have a lower potential of hydrogen or lower hydrogen solubility than the unstressed region. Furthermore, it is demonstrated that the Diffusion First Model (DFM) proposed by the author is defective in terms of kinetics because hydrogen diffusion before hydride nucleation just governs the rate of hydride nucleation, neither the rate of hydride growth nor the crack growth rate (CGR).     

Effect of unloading on crack growth rate of Zr-2.5Nb tubes

Crack growth rates (CGRs) of a heat-treated Zr-2.5Nb tube were determined using compact tension specimens with 60 ppm H at 250 °C under the constant and cyclic loads where the load ratio R was changed from 0.13 to 0.68. CGR was the highest under the constant load and decreased under the cyclic load with decreasing R despite a decrease of the critical hydride length indicating the enhanced rate of hydride cracking. Hence, the decreased CGR under the cyclic load is due to unloading during the cyclic load inducing the compressive stress at the crack tip. This compressive stress suppresses hydride nucleation rate, leading it to govern the CGR, according to Kim’s new model. Evidence is provided by citing Simpson’s experiment demonstrating that unloading from 15 MPa √m decreased the CGR of a cold-worked Zr-2.5Nb tube but annealing did the reverse. This study demonstrates for the first time that the retarded CGR due to an overload during the DHC tests is understood in view of crack growth kinetics using Kim’s model.     

Effect of radial hydrides on the axial and hoop mechanical properties of Zircaloy-4 cladding

The effect of radial hydrides on the mechanical properties of stress-relief annealed Zircaloy-4 cladding was studied. Specimens were firstly hydrided to different target hydrogen levels between 100 and 600 wt ppm and then thermally cycled in an autoclave under a constant hoop stress to form radial hydrides by a hydride reorientation process. The effect of radial hydrides on the axial properties of the cladding was insignificant. On the other hand, the cladding ductility measurements decreased as its radial hydride content increased when the specimen was tested in plane strain tension. A reference hydrogen concentration for radial hydrides in the cladding was defined for assessing the fuel cladding integrity based on a criterion of the tensile strength 600 MPa. The reference hydrogen concentration increased with the specimen (bulk) hydrogen concentration to a maximum of ∼90 wt ppm at the bulk concentration ∼300 wt ppm H and then decreased towards higher concentrations.     

Influence of metallurgical variables on delayed hydride cracking in Zr-Nb pressure tubes

During service, Zr-2.5Nb pressure tubes of nuclear power reactors may be prone to suffer from crack growth by delayed hydride cracking (DHC). For a given hydrogen plus deuterium concentration there is a critical temperature (T_C) below which DHC may occur. In this work, T_C was measured for specimens cut from pressure tubes made in Canada (CANDU) and in Russia (RBMK). Hydrogen was added to the specimens to get concentrations ranging from 24 to 60 wt ppm. It was found that T_C was higher than the corresponding precipitation temperature. The crack propagation velocity (V_P), measured in axial direction, increases from a minimum at T_C to a maximum at a temperature close but higher than the precipitation temperature. At lower temperatures, when hydride precipitates are present in the bulk, V_P follows an Arrhenius law: V_P = A exp(−Q/RT), with an activation energy Q of 66-68 kJ/mol for both tubes. The RBMK material presented lower velocities than CANDU one.     

CANDU及RBMK压力管锆合金的氢致延迟断裂研究

采用紧凑拉伸试样(CT），在恒定载荷、不同氢含量、不同温度条件下，测量了CANDU堆和RBMK堆Zr-2．5Nb压力管材料氢致延迟开裂速率。用金相显微镜和扫描电镜观察断口及氢化物形貌，并测量临界应力场强度因子及开裂速率，对材料的微结构及氢化物分布进行分析。结果表明，氢致延迟断裂(DHC)生长呈阶梯状。与CANDU压力管比较，RBMK压力管的DHC开裂速率将近低一个数量级。其原因是RBMK压力管的屈服强度比CANDU压力管低得多。     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.     

An elastoplastic phase-field model for the evolution of hydride precipitation in zirconium. Part II: Specimen with flaws

An elastoplastic phase-field model, described in Part I, was applied to bulk materials containing flaws such as sharp cracks and blunt notches. An additional set of long range order parameters, namely, stress-free strains for flaws, was introduced. The nucleation and growth of hydrides near a void or a crack were simulated by the proposed elastoplastic phase-field model. The effects of notch root radius, hydrogen concentration in solid solution, yield stress of the matrix and the level of externally applied stress on hydride morphology around flaws were studied. It is demonstrated that parameters such as the distribution of the tensile stress component perpendicular to the hydride platelet normal may be closely monitored during hydride growth near a flaw with or without externally applied stresses. Combined with a fracture criterion and real experimental data, the model is capable of predicting the rate and morphology of hydride precipitation, and crack initiation near flaws.     

Characterization of hydrogen concentration in Zircaloy-4 using ultrasonic techniques

The relationship between hydrogen concentration precipitated as hydride particles and ultrasonic parameters, such as velocity and attenuation, was examined in Zircaloy-4 samples for potential applications in the Non-Destructive Test Field. Different amounts of hydrogen (up to 517 ppm) were introduced in the samples by gaseous charging. Ultrasonic attenuation measurements were performed with compressive waves at frequencies of 10 and 30 MHz, and propagation velocity measurements were performed at 10 MHz. Ultrasonic velocity showed an approximately linear increase with hydrogen concentration and it could be used as an assessment parameter when the hydrogen level is high enough. Attenuation versus hydrogen concentration has been fitted by a logarithmic equation at 10 MHz. At 30 MHz a fluctuating behavior of the attenuation prevented measurement of the hydrogen concentration.     

Experimental studies of mechanical properties of solid zirconium hydrides

Zirconium alloy Zr-2.5Nb has been hydrided to ZrH_x (x = 1.15-2.0), and studied using microhardness and unconfined and confined compression techniques. At room temperature, results on Young’s modulus and yield strength of solid hydrides show that these mechanical properties remain about the same as the original zirconium alloy for hydrogen compositions up to about ZrH_1.5. The levels of these properties start to drop when δ hydride becomes the major phase and reaches a minimum for the ε hydride phase. Between room temperature and 300 °C, Young’s modulus of solid hydrides decreases with temperature at about the same rate as it does for the original zirconium alloy.     

含裂纹锆合金包壳管氢致多场耦合行为的数值模拟研究

考虑氢化物应力再取向,给出了锆合金包壳管氢致多场耦合行为的理论模型。建立了相应的多场耦合计算方法,编程获得了有限元程序。针对内压作用下的含轴向裂纹包壳管,建立了有限元模型,对其氢致多场耦合行为进行了计算分析。研究结果表明：对于含大量固溶氢原子的含裂纹包壳管,只有裂纹尖端区域析出较多的氢化物,这主要是由于此处存在很大的静水应力梯度和氢原子浓度梯度,并具有较低的氢原子固溶度;裂纹尖端析出的氢化物绝大部分沿包壳管径向,致使包壳管易于产生径向开裂,威胁其安全性;内压施加完成后,因氢化物析出膨胀,裂纹尖端区域的环向应力、径向应力、静水应力及其梯度均随时间而降低,导致氢化物析出逐渐减速。     

Identification and quantification of hydride phases in Zircaloy-4 cladding using synchrotron X-ray diffraction

Zirconium hydrides precipitate in fuel cladding alloys as a result of hydrogen uptake from the high-temperature corrosion environment of light water reactors. Synchrotron X-ray diffraction was performed at room temperature on stress-relieved Zircaloy-4 cladding with two distributions of hydrides - (1) uniformly distributed hydrides across the entire cladding wall and (2) hydride rim next to the outer surface. The δ-hydride phase was found to be the predominant hydride phase to precipitate for hydrogen contents up to 1250 weight parts per million (wt ppm). At a higher content, about 3000 wt ppm, although δ-hydride is still the majority phase, a significant amount of γ-hydride is also observed. At even higher hydrogen contents, in excess of approximately 6000 wt ppm, such as can occur in a highly dense hydride layer, peaks associated with the ε-hydride phase are also observed in the diffraction pattern. The volume fraction of hydrides was estimated as a function of hydrogen content using the integrated intensities of select diffraction peaks corresponding to the α-Zr matrix and the hydride phases. These estimated values agree well with calculated values from the independently measured concentrations. The results of this study indicate that hydride precipitation in Zircaloy-4 is a complex process of evolving hydride phases with increasing local hydrogen content.     

DSC “peak temperature” versus “maximum slope temperature” in determining TSSD temperature

One of the concerns of the nuclear industry is the deleterious effect of hydrogen on the structural integrity of the reactor core components due to delayed hydride cracking (DHC). The DHC process occurs when hydrogen concentration exceeds the terminal solid solubility (TSS) in the component. Thus, the accurate knowledge of TSS is necessary to predict the lifetime of the components. Differential scanning calorimetry (DSC) is normally used to measure the hydrogen TSS in zirconium alloys. There is a measurable change in the amount of heat absorbed by the specimen when the hydrides dissolve. The hydride dissolution process does not exhibit a well-defined “sharp” change in the heat-flow signal at the transition temperature. A typical DSC heat-flow curve for hydride dissolution has three definite features; “peak temperature” (PT), “maximum slope temperature” (MST) and “completion temperature”. The present investigation aims to identify the part of the heat-flow signal that closely corresponds to the TSS temperature for hydride dissolution (T_TSSD).Coupons were cut from a Zr-2.5Nb specimen, which had been previously hydrided using an electrolytic cell to create a surface hydride layer of ∼20 μm thick on all sides of the specimen. The coupons were then annealed isothermally at various temperatures to establish the T_TSSD under equilibrium conditions. Subsequently the hydride layer was removed and the coupons were analyzed for TSSD temperature using DSC. The PT and MST for each DSC run were determined and compared to the annealing temperature of the coupon. The results show that the annealing temperature (the equilibrium T_TSSD) is much closer to the DSC PT than any other feature of the heat-flow curve.     

The first step for delayed hydride cracking in zirconium alloys

Two models for delayed hydride cracking (DHC) in zirconium alloys are distinguished by their first step:

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The loading of a crack induces hydride precipitation. The hydride is postulated to create a hydrogen concentration gradient, where the bulk concentration is greater than that at the crack tip. This concentration gradient is taken as the driving force for diffusion of hydrogen to the crack tip, and subsequent hydride growth. This model is called the precipitate first model (PFM).

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The tensile stress at the crack tip induces a gradient in chemical potential that promotes the diffusion of hydrogen to the crack tip. Hydrides form if the hydrogen concentration reaches the solubility limit for hydride precipitation. The mechanism is postulated to create a hydrogen concentration gradient, where the bulk concentration is lower than that at the crack tip. The gradient in chemical potential is taken as the driving force for diffusion of hydrogen to the crack tip, and subsequent hydride growth. This model is called the diffusion first model (DFM).

The second model, DFM, is developed. This model is shown to describe the main features of the experimental observations of DHC, without invoking new phenomena, such as reduction in the solubility limit for precipitation of hydride, as required by the PFM.     

High-pressure hydriding of Zircaloy cladding by the thermogravimetry and tube-burst techniques

Hydriding kinetics of modified Zircaloy claddings was studied by the thermogravimetric method at 400 °C and the tube-burst technique at 315 °C. Some specimens were prefilmed with a thin oxide layer by air oxidation on both the inner and outer surfaces which were either pickled or blasted. In the thermogravimetric test, the hydriding rates of bare cladding specimens (no oxide prefilm) were in the range 0.9-1.6 mg/cm² h with little effect of the surface treatment. Incubation times were less than 1 h or even zero. In the tube-burst test, immediate and extensive hydrogen uptake was observed for these non-coated specimens. On the other hand, the cladding specimens with oxide prefilm exhibited lower hydriding rates ranging from 0.01 to 0.05 mg/cm² h and incubation times increased to 42 h. In addition, no hydrogen uptake was observed for all oxide-coated specimens for 100-750 h.     

Effect of hydride orientation on fracture toughness of Zircaloy-4 cladding

Hydrogen embrittlement is one of the major degradation mechanisms for high burnup fuel cladding during reactor service and spent fuel dry storage, which is related to the hydrogen concentration, morphology and orientation of zirconium hydrides. In this work, the J-integral values for X-specimens with different hydride orientations are measured to evaluate the fracture toughness of Zircaloy-4 (Zry-4) cladding. The toughness values for Zry-4 cladding with various percentages of radial hydrides are much smaller than those with circumferential hydrides only in the same hydrogen content level at 25 °C. The fractograghic features reveal that the crack path is influenced by the orientation of zirconium hydride. Moreover, the fracture toughness measurements for X-specimens at 300 °C are not sensitive to a variation in hydride orientation but to hydrogen concentration.     

Assessment of aging of Zr---2.5Nb pressure tubes in CANDU reactors

In modern CANDU^™ nuclear generating stations, pressure tubes of cold-worked Zr---2.5Nb material are used in the reactor core to contain the fuel bundles and the heavy water (D₂O) coolant. The pressure tubes operate at an internal pressure of about 10 MPa and temperatures ranging from about 250°C at the inlet to about 310°C at the outlet. Over the expected 30 year lifetime of these tubes they will be subjected to a total fluence of approximately 3 × 10²⁶ n m⁻². In addition, these tubes gradually pick up deuterium as a result of a slow corrosion process. When the hydrogen plus deuterium concentration in the tubes exceeds the hydrogen-deuterium solvus, the tubes are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC). If undetected, such a cracking mechanism could lead to unstable rupture of the pressure tube. A fitness-for-service methodology has been developed which assures that this will not happen. A key element in this methodology is the acquisition of data and understanding—from surveillance and accelerated aging testing—to assess and predict changes in the DHC initiation threshold, the DHC velocity and the fracture toughness (critical crack length) as a function of service time. The most recent results of the DHC and fracture toughness properties of CANDU pressure tubes as a function of time in service are presented and used to suggest procedures for mitigation and life extension of the pressure tubes.     

Hydride behavior in Zircaloy cladding tube during high-temperature transients

In order to study the hydride behavior in high burnup fuel cladding during temperature transients expected in anticipated operational occurrences and accidents, unirradiated hydrided Zircaloy-4 cladding tubes were rapidly heated to temperatures ranging from 673 to 1173 K and annealed for holding time ranging from 0 to 3600 s. Hydrides were localized in the peripheral region of the cladding tubes prior to the annealing, as observed in high burnup fuel cladding. The localized hydride layer (hydride rim) was annealed out, and the radial hydride distribution became uniform after the annealing at 873 K for 600 s, 973 K for 60 s, or 1173 K for 0 s. The annealing out of the hydride rim is caused by the phase transformation from the α + δ phase to the α + β or β phase in the hydride rim and the subsequent drastic increase in the solid solubility and diffusion of hydrogen in Zircaloy. On the other hand, the radial distribution and morphology of hydrides did not change at lower temperatures: Thus, the hydride remains almost intact below the phase transformation temperature for the short time ranges.     

The terminal solid solubility of hydrogen in irradiated Zircaloy-2 and microscopic modeling of hydride behavior

Differential scanning calorimetry (DSC) has been applied to elucidate the terminal solid solubility (TSS) of hydrogen in Zircaloy-2 cladding tubes and spacer bands irradiated in commercial BWRs. While recovery of irradiation defects during the first heating stage of as-irradiated specimens made the DSC peak of hydride dissolution dull or broader, no significant difference was detected in the TSS between unirradiated and irradiated Zircaloy-2, irrespective of fast neutron fluence. The effect of post-irradiation annealing on TSS was also examined. The results suggest almost no interaction between irradiation defects and dissolved hydrogen or hydrides at temperatures around 300 °C. Using the present TSS data and reported hydrogen- and hydride-related properties, a microscopic analysis code HYMAC for analyzing hydride behavior in cladding tube with textured grains was constructed. Stress-induced preferential precipitation and dissolution of hydrides were reproduced by adopting a TSS sub-model in which the solubilities decrease in proportion to stress normal to the habit plane in grains and to grain faces. Analyzed results by the code were consistent with typical experimental results of hydride behavior.     

Thermodynamic modeling of the Zr-H-Nb system on the Zr-rich side and application

A thermodynamic modeling of the Zr-H-Nb system on the Zr-rich side (wt.%Nb ? 2.5) was established using CALPHAD (CALculation of PHAse Diagrams) method. The calculated pressure-composition isotherms and the composition limits in the two-phase (bcc + fcc) region agree well with the experimental data. Meanwhile, some thermodynamic data of Zr-H-Nb system (wt.%Nb = 1, P_H2=101.325 kPa) was calculated, such as equilibrium hydrogen concentration, phase composition and phase transition temperature, which was applied to guide the preparation of the crack-free zirconium hydride. The hydriding of Zr-1 wt.%Nb alloys was carried out and the crack-free zirconium hydrides with hydrogen concentration of 1.6H/Zr (at.) and 1.8H/Zr were prepared respectively based on the calculated results.