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Influence of hydride precipitated at the tip of the crack on crack propagation in zircaloy-4 is numerically investigated using the extended finite element method (XFEM). Numerical simulation is performed on compact tension specimen to understand the effects of crack and hydride lengths on crack propagation in terms of stress intensity factor and J-integral. The values of stress intensity factor and J-integral indicate that hydride induces the crack instability. The stress intensity factor decreases with the precipitation of brittle hydride phase at the crack tip, resulting in hydride-assisted crack propagation. A comparison of crack propagation behaviour with different hydride lengths is also presented. The crack remains stable in the absence of the hydride while it propagates when hydride is considered at its tip for the same applied load. The crack arrests only after reaching to the zircaloy metal matrix causing complete fracture of hydride. [ABSTRACT FROM AUTHOR]

Although hydrogen uptake in Zirconium (Zr) based claddings has been a topic of many studies, hydrogen uptake as a function of alloy composition has received little attention. In this work, commercial Zr-based cladding alloys (Zircaloy-2, Zircaloy-4 and ZIRLO™), differing in composition but with similar initial textures, grain sizes, and surface roughness, were aqueously charged with hydrogen for 100, 300, and 1000 s at nominally 90 °C to produce hydride layers of varying thicknesses. Transmission electron microscope characterization following aqueous charging showed hydride phase and orientation relationship were identical in all three alloys. However, elastic recoil detection measurements confirmed that surface hydride layers in Zircaloy-2 and Zircaloy-4 were an order of magnitude thicker relative to ZIRLO™. [ABSTRACT FROM AUTHOR]

The pressurized heavy water reactors use natural uranium oxide (UO 2 ) as fuel and uses cladding material made up of zircaloy, an alloy of zirconium. Pickling of zircaloy tubes and appendages viz. , spacer and bearing pads is carried out to remove the oxide layer and surface contaminants, if present. Pickling solution, after use for many cycles i.e. , used pickling solution (UPS) is sold out to vendors, basically for its zirconium value. UPS, containing a relatively small concentration of hydrofluoric acid. After repeated use, pickling solution become saturated with zirconium fluoride complex and is treated by adding sodium nitrate to precipitate sodium hexafluro-zirconate. The remaining solution can be recycled after suitable makeup for further pickling use. The revenue lost by selling UPS is very high compared to its zirconium value, which causes monetary loss to the processing unit. Experiments were conducted to regenerate and reuse UPS which will save a good amount of revenue and also protect the environment. Experimental details and results are discussed in this paper. [ABSTRACT FROM AUTHOR]

Both the expected and an additional orientation relationship between α-Zr and δ-hydride in blistered zirconium alloys are explored through the reconstruction of the parent α-Zr phase from electron backscatter diffraction maps of δ-hydride. Parent and child variant relationships for the transformation are presented with the aim of reconstruction of the parent α-Zr grain structure and texture from orientation maps of the δ-hydride at varying distances from the blister centre in a recrystallized Zircaloy-4 sample. Up to 13% of the δ-hydride is found to be variants of the additional orientation relationship, the fraction of which decreases with increasing distance from the blister centre. Texture reconstructions by other experimental methods are accordingly suggested to incorporate the additional orientation relationship. [ABSTRACT FROM AUTHOR]

High-temperature oxidation of zirconium-based cladding materials is expected to be the primary cause of the fuel assemblies' degradation in spent fuel storage pool loss-of-cooling accidents. Here, high-temperature Zircaloy-4 oxidation has been studied through two-stage oxidation experiments using O isotope. Particular attention is paid to the effect of a low-temperature pre-oxidation scale which aims to simulate the corrosion scale existing on spent fuel cladding. Raman imaging was used to investigate the O distribution in the scales exposed either to O or to O + N atmosphere at high temperature. Results were assessed against more conventional SIMS mapping. The use of the O isotope tracer technique associated with micro-Raman mapping of the scales is demonstrated to be a powerful method to investigate the transport properties in the scales and will help to gain understanding of the kinetic differences between different pre-oxides. [ABSTRACT FROM AUTHOR]

This paper presents the methodology used for manufacturing the Zircaloy-4 tubes required for the Zircaloy-4 mock-up dissolver assembly. The evolution of the microstructure at different stages of production of Zircaloy-4 tubes was characterized using scanning electron microscopy (SEM). Microstructural evolution follows the sequence as dendritic structure (as-cast ingot) → Widmanstätten structure (β-quenched) → bimodal grain size (hot extruded) → heterogeneous deformed structure (pilgered) → partially recrystallized structure (final annealed). High strength and low ductility were obtained for the Zircaloy-4 tubes at pilgered condition due to grain size refinement and work hardening during the pilgering process. Compared with the pilgering stage, the Zircaloy-4 tubes in the final annealed condition exhibited moderate strength and high ductility due to the partially recrystallized microstructure. Autoclaving was carried out to improve the corrosion properties of the pilgered Zircaloy-4 tubes in boiling 11.5 M nitric acid. When exposed to boiling 11.5 M nitric acid for 1000 h, a lowest corrosion rate of 0.003 mpy was obtained for autoclaved Zircaloy-4 tubes. Laser Raman Spectroscopy (LRS) analysis confirmed that the origin of passivity of pilgered and autoclaved Zircaloy-4 tubes was due to the presence of a protective passive film composed of ZrO2. [ABSTRACT FROM AUTHOR]

Ultrasonication is proposed to accelerate the growth of coating formed on a zircaloy-4 by plasma electrolytic oxidation (PEO) at low electric power. PEO was conducted at a DC current density of 20 mA/cm2 and a limiting voltage of 300 V for 5–15 min. The PEO bath contained a balanced ratio of sodium phosphate, sodium silicate, and potassium hydroxide. Surface investigation using optical and electron microscopes on the resulting coatings revealed a significant enhancement of Faradaic efficiency from 15–17 % to 38–46 % due to sonication. Ultrasonication-assisted PEO (UPEO) accelerated the coating growth by lowering the breakdown voltage, generating more plasma discharge, and avoiding the gas shielding effect. Consequently, the sonication did not alter the structure and composition of the coatings. The coatings were composed of amorphous SiO 2 as the matrix and reinforced by the crystalline ZrO 2 , as confirmed by the x-ray diffraction (XRD) and energy dispersive x-ray spectroscopy (EDS) analyses. The combined effect of composition, porosity, and thickness yielded an insignificant difference in the hardness and wear-rate, and the corrosion resistance between the PEO and UPEO coatings. Electrochemical measurements demonstrated an order magnitude enhancement of corrosion resistance as a result of coating. The inner layer served effectively as the corrosion barrier, while the mediate and outer layers gave significant protection against the mechanical load. The faster growth rate provided by ultrasonication is beneficial for the extensive development of PEO technology. [Display omitted] • PEO coating was performed on Zr-4 using a low current density of 20 mA/cm2. • The coating growth rate was accelerated 2–7 times by ultrasonication. • The coating was composed of an SiO 2 inner layer and an SiO 2 /ZrO 2 outer layer. • The ultrasonication enhanced the formation of tetragonal and monoclinic ZrO 2 phases. • The mechanical and corrosion properties of PEO coating were preserved. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Zirconium and zirconium alloys are widely used as nuclear reactor core materials such as fuel cladding and guide tubes because they have excellent corrosion- and radiation-resistant properties. In the reactor core, zirconium alloys are subjected to high-energy neutron fluence, causing radiation-induced dislocation and growth. To discern a possible correlation between radiation-induced dislocation and growth, characteristics of dislocation and growth in zirconium and its alloys are examined. The radiation-induced dislocation including and dislocation loops is reviewed in various temperature and fluence ranges, and their growth behavior is examined in the same way. To further a fundamental understanding, radiation-induced growth prediction models are briefly reviewed. This research will assist in the design of zirconium based components as well as the safety analysis of various reactor conditions, in both research and commercial reactors.

In order to study the relative effectiveness of oil-based and graphene-based lubrications towards suppressing the friction and wear damage of zircaloy-4 during low amplitude linear reciprocatory wear, fretting wear tests were performed with varying normal loads (5–20 N) under unlubricated, oil-lubricated (using Castrol TDN-81) and graphene-lubricated/coated (using ∼7 layers graphene film) conditions against stainless steel ball. While the oil-based lubricant has been found to be more effective towards lowering the co-efficient of friction at lower loads, the effectiveness of the graphene-based lubricant/coating increase significantly with an increase in load. In the contexts of suppressing the extent/severity of wear damage, the graphene-based lubricant/coating has been found to be significantly more effective than the oil-based lubricant. In fact, increase in effectiveness of the graphene-based lubricant/coating with load results in the wear volume remaining nearly invariant upon increase in load; thus, causing reduction in wear rate. This is in contrast to the increase in wear volume/rate with increase in load for the unlubricated and oil lubricated conditions; such that the wear volume and rate are lower for the graphene-lubricated/coated condition by more than a factor of 2 with respect to the oil-lubricated condition at 20 N. In the unlubricated condition severe sub-surface cracking induced material removal/delamination, presumably aided by incorporation of Fe and O due to tribochemical reaction, form the major mechanisms of wear damage of zircaloy-4. By contrast, all these get suppressed in the presence of the graphene-based lubricant/coating, even at 20 N; thus, leading to significant reduction in the overall wear damage. [Display omitted] • Graphene is more effective than oil, as lubricant, towards supressing material removal during fretting wear of zircaloy-4. • Effectiveness of graphene-based lubricant increases with increase in normal load, in contrast to that of oil lubricant. • Graphene-based lubricant supresses material transfer from counterbody, tribochemical wear and sub-surface cracking. • Differential roles and importance of oil and graphene-based lubricants towards supressing friction/wear damage of zircaloy. [ABSTRACT FROM AUTHOR]

During operation in a nuclear reactor, Zr-based nuclear fuel cladding is subject to waterside corrosion which can lead to hydrogen ingress. The hydrogen that enters the material will migrate to colder spots and precipitate as zirconium hydrides if the hydrogen content exceeds the hydrogen terminal solid solubility in the material. Since a temperature gradient is established in the radial direction of the cladding during operation, the hydrides can preferentially precipitate at the colder outer surface of the cladding. Other gradients can also occur in the longitudinal and azimuthal directions of the cladding tube. As a consequence, hydrogen redistributes itself in response to the concentration and temperature gradients present in the sample. The response of the hydrogen in solid solution to temperature gradients is governed by the heat of transport Q* as a function of temperature, so it can be used in the BISON code. A set of experiments was set up to determine the heat of transport (Q*), in which a uniformly hydrided Zircaloy-4 sample is annealed under a fixed temperature gradient at a range of temperatures, and the resulting hydrogen distribution is analyzed to determine Q*. The results are discussed in terms of existing literature. [ABSTRACT FROM AUTHOR]

The development and verification of nuclear fuel-cladding performance codes require extensive testing to establish empirical correlations, necessitating more accelerated testing methods that can replace large validation studies. In the present work a technique was developed and implemented to experimentally quantify the relationship between temperature, cladding internal pressure, and strain in-situ during a burst test, with the specific aim of generating data for code validation and model refinement. Digital image correlation was used to measure hoop, axial, and radial strain, and infrared thermography to quantify axial temperature gradients. Several key experimental modifications to traditional LOCA burst testing were necessary, but are shown to effectively have little impact on burst temperatures. The time/temperature dependent pressure and time/spatially dependent thermal gradient data were used with BISON to simulate cladding burst and strain rates. Although the measured DIC strains reach appreciable amounts at slightly lower temperatures than BISON simulated strains, there is good agreement between the measured and simulated strains in terms of magnitude and rates leading up to burst. However, in the few seconds prior to and during burst, the BISON simulated strains are significantly lower than the measured strains, indicating the potential for model improvement. The combined experiment and simulation technique may be applied to any new cladding concept to accelerate fuel qualification. [ABSTRACT FROM AUTHOR]

Hydride reorientation experiments for unirradiated Zr-Nb alloy cladding tubes under multi-axial stress states induced by internal pressurization were performed with peak cladding temperature set to 400 °C. Applied internal pressure was held constant during cooling down stage. PROPHET was used to analyze Radial Hydride Fraction (RHF) of tested cladding tubes. The attained RHF behavior for a wide range of rod internal pressure (7.5-18.5 MPa) and hydrogen content (0-1100 wppm) were investigated. Empirical models for RHF with respect to rod internal pressure and hydride contents were developed. Ring compression tests (RCTs) were conducted to investigate toughness (strain energy density) change with radial hydride formation. In order to take the multi-axial stress effect into account, rod internal pressure-based regulation is necessary. Hydride reorientation threshold rod internal pressure is ∼9.32 MPa. The rod internal pressure corresponding to hoop stress of 90 MPa is 11.5 MPa, indicating that hydride reorientation in multi-axially stressed Zircaloy cladding tube occurs at hoop stresses below 90 MPa. The presented thermodynamic model for threshold rod internal pressure gives improved agreement with experimentally measured threshold with input the directly measured orientation relationship between δ-hydride and α -Zr. [ABSTRACT FROM AUTHOR]

Samples of Zircaloy-4 were ion irradiated at several doses and irradiation temperatures and studied using transmission electron microscopy to discern the amorphization and dissolution behavior of second-phase precipitates under irradiation. It is found that near the critical temperature for amorphization under neutron irradiation a similar amorphization morphology is obtained under proton irradiation as under neutron irradiation, that is, an amorphous layer starting at the precipitate-matrix interface and moving in with increasing dose. However, the rate of amorphous layer advancement is much slower than that seen under neutron irradiation, and saturates with dose so that it remains a partial precipitate amorphization, i.e., this precipitate amorphization mechanism does not lead to complete amorphization as it occurs with neutrons. The results indicate that the critical temperature for bulk amorphization is primarily dependent on the displacement cascade density achievable with the irradiating particle and secondarily with dose rate. Depletion of iron from the precipitates is also observed, although not directly linked to amorphization. [ABSTRACT FROM AUTHOR]

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 ( 14 C), which is mainly generated by the 14 N(n,p) 14 C 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 14 C 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 14 C specific activity of the base metal was 1.49×10 4 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 17 O(n,α) 14 C reaction, the 14 C 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, 14 C 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 14 C 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 14 C was from the external oxide layer and 3.3% was from the base Zircaloy metal. However, both the 14 C abundance and the low leaching rate suggests that 14 C in oxide does not have a significant impact on the IRF in the safety case. [ABSTRACT FROM AUTHOR]

For a better understanding of the decommissioning of the Fukushima-daiichi nuclear power plant, the melting behavior of the control blade and the channel box should be clarified. In Fukushima nuclear reactor, the channel box was made of Zircaloy-4, and the control rode is made of B 4 C together with stainless steel cladding and sheath. In the study, the interaction among B 4 C, stainless steel (SUS), and Zircaloy-4 was investigated at 1473 K in either argon or air atmosphere. In argon, Zircaloy is melted by the diffusion of elements from SUS, and SUS was melted at 1473 K by the diffusion of C and B. In air, SUS reacted with B 2 O 3 and formed an oxides melt firstly. Then, the oxidized Zircaloy contacted with this melt and fused. Moreover, the progress of core melting process during severe accident under different atmospheres was firstly discussed. [ABSTRACT FROM AUTHOR]

A hot vacuum extraction-isotope dilution mass spectrometry (HVE-IDMS) was studied for determination of hydrogen isotopes in zirconium metal and alloys as nuclear reactor materials. A theoretical assessment of the completeness of the extraction of hydrogen isotopes under the chosen condition was carried out based on the hydrogen and deuterium solubility data for zirconium. The optimal isotopic spiking condition for conventional IDMS was further explored for the special case IDMS where the isotope abundance of the samples is varied and non-natural. Applying the optimal conditions, the accurate IDMS determination was realized. The agreement between the measured values and the certified or prepared values of standard reference materials and homemade standard materials validate the method developed. [ABSTRACT FROM AUTHOR]

Zircaloys, rooted in their inherent passivity, are widely used as the cladding materials for light water reactors (LWRs) due to their desirable corrosion resistance against high temperature and high pressure water. The hydrogen permeation behavior on zircaloy with passive film, however, is rarely understood so far. In this work, the gaseous hydrogen permeation characteristic mainly on the defect evolution of nanoscale passive films of nanoscale passive films formed hydrothermally on different types of zirconium alloys (Zirconium, Zircaloy-4, N18 and M5) was explored and compared. Surface analytical techniques were implemented to evaluate the hydrogen permeation characteristic. The AES depth profiles showed that the as-prepared oxides were hundreds of nanometers in thickness and inward diffusion occurred upon hydrogen exposure. Electrochemical measurements suggested that the oxides formed on different types of zircaloys were of similar phase structures of monoclinic zirconia but with different oxygen vacancy concentrations. XPS results comparison confirmed that niobium enhances the hydrogen resistance of the oxides, yet tin causes degradation of the oxides in the reductive environment. In view of these analyses, the involving mechanism was first proposed and discussed on the basis of point defect reactions. [ABSTRACT FROM AUTHOR]

Several specimens of Zircaloy-4 were creep tested at a single stress-temperature condition, and interrupted at different accumulated strain levels. Substructural observations were performed using bright field scanning transmission electron microscopy (BF STEM). The dislocation substructure was characterized to ascertain how creep strain evolution impacts the Modified Jogged-Screw (MJS) model, which has previously been utilized to predict steady-state strain rates in Zircaloy-4. Special attention was paid to the evolution of individual model parameters with increasing strain. Results of model parameter measurements are reported and discussed, along with possible extensions to the MJS model. [ABSTRACT FROM AUTHOR]

Aqueous reprocessing of plutonium-rich mixed oxide fuels require fluoride as a dissolution catalyst in boiling nitric acid for an effective dissolution of the spent fuel. High corrosion rates were obtained for the candidate dissolver materials zircaloy-4 (Zr-4) and commercial pure titanium (CP-Ti grade 2) in boiling 11.5 M HNO 3 + 0.05 M NaF. Complexing the fluoride ions either with Al(NO 3 ) 3 or ZrO(NO 3 ) 2 aided in decreasing the corrosion rates of Zr-4 and CP-Ti. From the obtained corrosion rates it is concluded that CP-Ti is a better dissolver material than Zr-4 for extended service life in boiling 11.5 M HNO 3 + 0.05 M NaF, when complexed with 0.15 M ZrO(NO 3 ) 2 . XPS analysis confirmed the presence of TiO 2 and absence of fluoride on the surface of CP-Ti samples, indicating that effective complexation had occurred in solution leading to passivation of the metal and imparting high corrosion resistance. [ABSTRACT FROM AUTHOR]

Zircaloy is a kind of cladding material for monolithic nuclear fuel plates, which has a strong strain-rate-dependent mechanical constitutive relation at the rolling temperature. In order to model the rolling process of monolithic nuclear fuel plates, it is critical to effectively define the mechanical properties of Zircaloy in the commercial finite element software. Different three-dimensional stress-updating algorithms are built, respectively, including a fully implicit algorithm, a simplified implicit one and an explicit one. The corresponding VUMAT subroutines suitable for explicit dynamic rolling simulation for a Zircaloy plate are developed, and the computation precision and efficiency are compared in ABAQUS. The simplified implicit stress-updating algorithm is determined to be the best choice. The effects of temperature, rolling velocity and reduction on the strain rate distribution, the deformation and rolling force are investigated. [ABSTRACT FROM AUTHOR]

This paper presents detailed analyses of a pressurized water reactor with a new reflector design using zirconium metal. The optimization of the reflector design has been performed using a two-dimensional fuel assembly reflector model. The three-dimensional core calculation results with the optimized reflector were compared against those with the existing water reflector and iron reflector. The high scattering cross section of zirconium enhances neutron reflections from the reflector to the core, increasing the peripheral assembly powers. From the analysis based on the equilibrium core, it was noted that the cycle length can be extended, and the pin peaks can be decreased when using zirconium reflector. The analysis has been performed for the optimized power reactor 1000 core with combustion engineering type fuel assemblies using the CASMO-4E/SIMULATE-3 (Studsvik Scandpower, Inc., Waltham, MA, USA) code system and SERPENT (VTT Technical Research Centre of Finland, Vuorimiehentie 3, 02150 Espoo, Finland) code, with ENDF/B-VI data. Copyright © 2015 John Wiley & Sons, Ltd. [ABSTRACT FROM AUTHOR]

Zircaloy-4 was tested at room-temperature over a range of hydrogen content between 10 and 200 ppm, and stress-states between a triaxiality of −0.23 and 0.9. Triaxiality ( η ) is defined as the ratio of hydrostatic stress to von Mises stress and was controlled through use of select mechanical test specimen designs. Testing of smooth and notched tensile specimens ( η = 0.33 to 0.9) results in an increase in the stress to initiate plastic deformation and a decrease in strain to failure at higher values of η . Increases in triaxiality are shown to have a more significant effect on reducing the strain to failure when the material is hydrided. Increases in strain to failure are observed at lower values of triaxiality for dual keyhole specimens ( η = 0.1) and compression specimens ( η = −0.17 to −0.23), with hydrided material showing much less decrement in strain to failure at these lower triaxialities. Examinations of microstructure are used to show that a change in mechanism for deformation and fracture with triaxiality can explain these differences in mechanical behavior and a model is developed to describe the observed changes in strain to failure with stress-state. [ABSTRACT FROM AUTHOR]