Observation of crack microstructure in oxides and its correlation to oxidation and hydrogen-uptake by 3D FIB Tomography – case of Zr-ZrO2 in reactor

Abstract

Zirconium based alloys are used as fuel claddings in Light Water Reactors due to their good resistance to degradation and low neutron absorption cross section. However, life limiting processes occur during the service of the cladding such as oxidation and hydrogen-uptake. During the oxidation of the material, hydrogen enters the metal and it precipitates as brittle hydrides. In this study the 3D microstructure of a high burn-up and a low-burnup LK3/L Zircaloy-2 cladding is characterized and compared using FIB Tomography. 3D reconstruction of the oxides of the claddings shows that the crack volume fraction increases with the number of cycles in the reactor, reducing its protectiveness against further corrosion and H-uptake. The visualization of the metal-oxide interface revealed that the oxidation of the hydrides in the metal could induce crack formation in the oxide and therefore it could be one of the causes of the increasing oxidation and H-uptake in this material.     

Behavior and failure of uniformly hydrided Zircaloy-4 fuel claddings between 25 °C and 480 °C under various stress states,including RIA loading conditions

The anisotropic plastic behavior and the fracture of as-received and hydrided Cold-Worked Stress Relieved Zircaloy-4 cladding tubes are investigated under thermal–mechanical loading conditions representative of Pellet–Clad Mechanical Interaction during Reactivity Initiated Accidents in Pressurized Water Reactors. In order to study the combined effects of temperature, hydrogen content, loading direction and stress state, Axial Tensile, Hoop Tensile, Expansion Due to Compression and hoop Plane Strain Tensile tests are performed at room temperature, 350 °C and 480 °C on the material containing various hydrogen contents up to 1200 wt. ppm (hydrides are circumferential and homogeneously distributed). These tests are combined with digital image correlation and metallographic and fractographic observations at different scales. The flow stress of the material decreases with increasing temperature. The material is either strengthened or softened by hydrogen depending on temperature and hydrogen content. Plastic anisotropy depends on temperature but not on hydrogen content. The ductility of the material decreases with increasing hydrogen content at room temperature due to damage nucleation by hydride cracking. The plastic strain that leads to hydride fracture at room temperature decreases with increasing hydrogen content. The influence of stress triaxiality on hydride cracking is negligible in the studied range. The influence of hydrogen on material ductility is negligible at 350 °C and 480 °C since hydrides do not crack at these temperatures. The ductility of the material increases with increasing temperature. The evolution of material ductility is associated with a change in both the macroscopic fracture mode of the specimens and the microscopic failure mechanisms.     

Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for β-Zr.Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.     

Investigation of failure behavior of two different types of Zircaloy clad tubes used as nuclear reactor fuel pins

Nuclear reactor fuel pins act as barriers to the release of radioactive fission products to the coolant flowing around these thin-walled tubes and hence they prevent the leakage of radioactivity to the surroundings of reactor core. These tubes are of small thickness in order to have less resistance in the path of heat flow from the fuel to the coolant. Investigation of failure behavior of these fuel clad tubes is of utmost importance to the designers and plant operators in order to ensure the maximum residence time of the fuel bundles inside the reactor core as well as to ensure minimal activity during operation and refueling activities. Various types of zirconium based alloys are used to manufacture these pins. The focus is to obtain better strength, ductility, corrosion resistance, oxidation resistance, and minimal creep including those due to irradiation-assisted damage and deformation processes. Two number of such types of alloys, namely, re-crystallization annealed Zircaloy-2 and stress-relief annealed Zircaloy-4, have been investigated in this work for their fracture behavior. As standard fracture mechanics specimens cannot be machined from these thin-walled tubes, non-standard specimens with axial cracks have been used in this work. Load normalization technique has been used to evaluate crack growth during loading of these specimens. It was observed that the re-crystallization annealed Zircaloy-2 specimens have higher initiation fracture toughness as well as higher resistance to crack growth compared to the other type of specimens. In order to understand the micro-structural aspects of the fracture resistance behavior of these materials, further investigation incorporating optical and transmission electron microscopy have also been carried out. It was concluded that the higher fracture resistance behavior of the re-crystallization annealed Zircaloy-2 specimens can be attributed to the presence of finer grain and sub-grain micro-structure, very low dislocation density and other defects in the material.     

Application of neutron radiography to study material processes during hypothetical severe accidents in nuclear reactors

Neutron radiography was applied for investigations of nuclear fuel cladding and control rod behaviour during steam oxidation at temperatures between 1123 and 1673?K under severe nuclear accident conditions. This article gives an overview of these investigations. At KIT, loss of coolant and severe nuclear accidents were experimentally simulated. Post-test examinations of damaged control rods were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. It could be seen clearly from these data that local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube. The uptake of hydrogen during steam oxidation and its diffusion in Zircaloy-4 was investigated in ex situ and in situ radiography experiments at temperatures of 1123 to 1673?K. The kinetics of hydrogen uptake and diffusion was determined. The oxide layer morphology strongly influences the hydrogen concentration in steam oxidized zirconium alloys. Differences of nearly one order of magnitude were found in samples withdrawn from large scale QUENCH experiments. The hydrogen diffusion coefficients were determined for various temperatures. Whereas the diffusion coefficients at 1123 and 1173?K agree well with values expected from literature values for pure Zr, at higher temperatures a faster diffusion was found. The determined activation energy of the hydrogen diffusion is about 10?% higher than published values in the literature.     

3D fuel cracking modelling in pellet cladding mechanical interaction

This study is concerned with the modelling of fuel behaviour and of pellet–cladding interaction (PCI). A new fuel software (PLEIADES) is currently co-developed by the Atomic Energy Commission (CEA) and Electricité de France (EDF). This software includes a multi-dimensional FE program (ALCYONE) devoted to Pressure Water Reactors (PWR) fuel rods. PCI studies are mainly undertaken with the 3D model of ALCYONE. The objectives of this work are twofold: first, to propose a constitutive model for the fuel pellet which accounts for the stress relaxation of the material resulting from cracking and creep, second, to estimate the impact of the pellet cracking on PCI.

In this paper, a mathematical formulation which couples a viscoplastic law for creep with a multi-surface plastic softening law for cracking is detailed, leading a two inelastic strains model. Mesh dependency is overcome thanks to a material parameter related to the finite element size. The 3D calculations of PCI presented in this paper show that the considered modelling of fuel cracking is consistent with the experimental knowledge available on crack development under irradiation. A parametric study is then presented which leads to the conclusion that the tangential stresses at the pellet cladding interface and hence the risk of PCI failure are significantly reduced when the fuel tensile strength is divided by two.     

Effect of hydrogen on the low-temperature mechanical properties of V-5 at % Ti alloys

The effects of grain size and hydrogen in solid solution or as hydrides on the strength and ductility of V-5 at % Ti was studied over the temperature range 15–448 K. Comparison of the strength and ductility characteristics of hydrogenated alloys where hydrides were not observed down to 78 K (1.8 and 1.9 at % H alloys) or where hydrides were observed to form near 230 K (3.8 and 3.9 at % H alloys) indicated that the presence of hydride precipitates had no apparent influence on the strength or ductility characteristics. It appears that the main consequence of hydride precipitation is that hydrogen is removed from solid solution making strengthening less effective than expected based on the total hydrogen content. Decreasing grain size from 31 m to 8 m had no apparent effect on ductility in the nonhydrogenated alloys (< 0.05 at % H) but it did increase the strength over most of the temperature range and especially at 15 K. In the hydrogenated alloys this decrease in grain size lowered the transition temperature about 10 K and it appreciably increased the degree of ductility return at 78 K and below. The ductility return below 78 K peaked near 50 K before decreasing below 30 K with the improvement in ductility being greatest in the alloys with the lower hydrogen contents.     

Investigation of Backscatter X-Ray Imaging Techniques for Uranium Dioxide Fuel Rods

Radiography by selective detection (RSD), was investigated for its ability to determine the presence and types of defects in a UO₂ fuel rod surrounded by zirconium cladding. Images created using a Monte Carlo model compared favorably with actual X-ray backscatter images from mock fuel rods. A fuel rod was modeled as a rectangular parallelepiped with zirconium cladding, and pencil beam X-ray sources of 160 kVp (79 keV avg) and 480 kVp (218 keV avg) were generated using the Monte Carlo N-Particle Transport Code (MCNP) to attempt to image void and palladium (Pd) defects in the interior and on the surface of the fuel pellet. It was found that the 160 kVp spectrum was unable to detect the presence of interior defects, whereas the 480 kVp spectrum detected them with both the standard and the RSD backscatter methods, though the RSD method was very inefficient. It was also found that both energy spectra were able to detect void and Pd defects on the surface using both imaging methods. Additionally, two mock fuel rods were imaged using a backscatter X-ray imaging system, one consisting of hafnium pellets in a Zircaloy-4 cladding and the other consisting of steel pellets in a Zircalloy-4 cladding which was then encased in a steel cladding (a double encapsulation configuration employed in irradiation and experiments). It was found that the system was capable of detecting individual HfO₂ pellets in a Zircaloy-4 cladding and may be capable of detecting individual steel pellets in the double-encapsulated sample. It is expected that the system would also be capable of detecting individual UO₂ pellets in a Zircaloy-4 cladding, though no UO₂ fuel rod was available for imaging.     

乏燃料容器设计中组件包壳的氢脆特性影响分析

目的针对高燃耗乏燃料在干法运输条件下存在的包壳脆性变化,探讨其对我国现阶段乏燃料运输及容器安全研究中以“组件结构保持完整”作为设计基准所带来的影响和考虑。方法结合锆包壳氢化物韧脆转变的机理,针对乏燃料离堆和运输过程对包壳性能变化进行分析,并根据美国“ISG-11”等技术导则中提出的判别准则,探讨事故载荷下高燃耗乏燃料包壳结构性能的变化。结果通常认为2G乏燃料不需要考虑其材料氢脆影响问题,而高燃耗的3G乏燃料则必须综合评价离堆及后续干法运输过程中各种因素变化对其性能的影响,以判断包壳的峰值温度、韧脆转变温度和温度变化幅度等是否会对事故工况下“其结构始终保持完整”的设计基准造成影响。结论随着我国进入3G高燃耗乏燃料密集运输的时代,在乏燃料运输容器设计及运输安全分析时更应充分考虑包壳材料氢脆特性影响下的乏燃料结构在事故载荷下的保持能力。     

Structure of thermally desorbed CeNi<Subscript>3</Subscript>-based hydrides

CeNi₃H _x (x = 0.7, 0.8, 1.0, 1.8, 3.4, 3.8) hydrides have been prepared through hydrogen desorption from CeNi₃ hydrogenated at low (p _{H 2} = 0.01 GPa) and high (p _{H 2} = 0.2 GPa) hydrogen pressures. Using X-ray and neutron diffraction, the hydrides are shown to be isostructural with CeNi₃ (sp. gr. P6₃/mmc, no. 194). The lattice parameters of the hydrides vary appreciably with hydrogen content. The sequence of hydrogen release from different interstices in the desorption process is shown to be opposite to that of hydrogen uptake in the hydrogenation process. The solid-solution range in the desorbed hydrides is much broader than that upon hydrogenation. The extent of the solid solution is influenced by the phase composition of the parent intermetallic compound.     

Improving the Hydrogen Reaction Kinetics of Complex Hydrides

Alanates, borohydrides, and amides are complex hydrides with high concentration hydrogen that have been actively investigated for materials‐based hydrogen storage on‐board polymer electrolyte membrane fuel cell (PEMFC) vehicle applications. The major challenge is to release hydrogen at fuel cell working temperature range at fast enough rate without simultaneous desorption of fuel cell poisoning impurities. We review recent progress in hydrogen reaction mechanism and schemes for complex hydride hydrogen storage.     

储氢材料的研究进展

氢能源以其可再生性和良好的环保效应成为未来最具发展潜力的能源载体,氢能被公认为人类未来的理想能源,而氢的储存是发展氢能技术的难点之一.本文综述了目前主要的储氢材料,如合金储氢、配位氢化物储氢、碳质材料储氢、有机液体氢化物储氢,并对未来的储氢材料发展进行了展望.     

轻质金属-铝氢化物贮氢材料的研究进展

介绍了几种轻质金属-铝氢化物贮氢材料的吸放氢机理和研究进展.轻质金属-铝氢化物贮氢密度高,但存在动力学性能差、放氢温度高、可逆反应程度低等缺点,目前主要通过掺杂催化剂、降低材料的颗粒尺寸等方法来创提高吸放氢的速率和效率.随着研究的深入,轻质金属-铝氢化物在贮氢方面将有广阔的发展前景.     

钒基固溶体型贮氢材料的研究进展

氢作为一种新能源,其制备、贮存技术有了迅速的发展.介绍和评述了钒基固溶体型贮氢材料的制备方法;综述了钒基固溶体型合金的组织结构、贮氢性能和电化学性能以及它们之间的关系;指出了今后钒基固溶体型贮氢材料应用研究的重点.     

Modeling of structural failure of Zircaloy claddings induced by multiple hydride cracks

Zirconium alloys have been serving as primary structural materials for nuclear fuel claddings. Structural failure analysis under extreme conditions is critical to the assessment of the performance and safety of nuclear fuel claddings. This work focuses on simulating structural failure of Zircaloy tubes with multiple hydride defects through modeling explicit crack propagation in ductile media. First, we developed an integrated cladding failure model by taking into account both crack initiation induced by hydride/matrix interface separation and ligament tearing-off between activated hydride cracks. Second, to accommodate the initiation, propagation, and coalescence of multiple cracks in finite plastic media we incorporated this structural failure model into a coupled continuous/discontinuous Galerkin (DG) based finite element code, a traditionally preferred implicit numerical framework. Third, to improve the adaptive placement of DG interface elements for crack propagation and to identify potential coalescence of cracks due to the interaction between adjacent hydride cracks, we defined a special failure index for the assessment of potential failure zones using both true plastic strain developed and predicted failure strain based on the Johnson–Cook material failure criterion. Finally, by calibrating the proposed material failure model using a cluster of Zircaloy material experimental tests, we successfully simulated a complete failure process of a fuel cladding tube with multiple hydride cracks.     

Simulation of the coupling behaviors of particle and matrix irradiation swelling and cladding irradiation growth of plate-type dispersion nuclear fuel elements

Within plate-type dispersion nuclear fuel elements, besides irradiation swelling of fuel particles induced by nuclear fissions, the metal matrix and the cladding are attacked continuously by the fast neutrons released from the fuel particles. As a consequence, the matrix undergoes a bit irradiation swelling and the cladding takes on irradiation growth, which both might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, the three-dimensional large-strain constitutive relations for the fuel particles, the metal matrix and cladding are developed; based on them, the method of virtual temperature increase proposed by Ding et al. (2008) is further developed to model the irradiation swelling; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-temperature-loadstep is proposed to simulate the coupling features of the irradiation swellings of both the metal matrix and the fuel particles together with the irradiation growth of the cladding. In order to clarify the critical factors that affect their mechanical performances and carry out optimal design, with the aid of the research thoughts of particle-reinforced composites, numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate the effects of irradiation swelling of the matrix and irradiation growth of the cladding as that: (1) they might weaken the in-pile mechanical performances at the matrix to some extent; and (2) the former increases interfacial stresses between the fuel meat and the cladding, while the latter relatively relieve those interfacial stresses; and the interfacial mechanical strength might be improved by getting suitable irradiation growth mode of the cladding.     

Recent advances and remaining challenges of nanostructured materials for hydrogen storage applications

The rapid and extensive development of advanced nanostructures and nanotechnologies has driven a correspondingly rapid growth of research that presents enormous potential for fulfilling the practical requirements of solid state hydrogen storage applications. This article reviews the most recent progress in the development of nanostructured materials for hydrogen storage technology, demonstrating that nanostructures provide a pronounced benefit to applications involving molecular hydrogen storage, chemical hydrogen storage, and as supports for the nanoconfinement of various hydrides. To further optimize hydrogen storage performance, we emphasize the desirability of exploring and developing nanoporous materials with ultrahigh surface areas and the advantageous incorporation of metals and functionalities, nanostructured hydrides with excellent mechanic stabilities and rigid main construction, and nanostructured supports comprised of lightweight components and enhanced hydride loading capacities. In addition to highlighting the conspicuous advantages of nanostructured materials in the field of hydrogen storage, we also discuss the remaining challenges and the directions of emerging research for these materials.     

高容量储氢材料的研究进展

氢能是一种理想的二次能源.氢能开发和利用需要解决氢的制取、储存和利用3个问题,而氢的规模储运是现阶段氢能应用的瓶颈.氢的储存方法有高压气态储存、低温液态储存和固态储存等3种.固态储氢材料储氢是通过化学反应或物理吸附将氢气储存于固态材料中,其能量密度高且安全性好,被认为是最有发展前景的一种氖气储存方式.由轻元素构成的轻质高容量储氢材料,如硼氢化物、铝氢化物、氨摹氢化物等,理论储氢容量均达到5%(质量分数)以上,这为固态储氢材料与技术的突破带来了希望.新型储氢材料未来研究的重点将集中于高储氢容量、近室温操作、可控吸/放氢、长寿命的轻金属基氢化物材料与体系.     

高容量储氢材料的研究进展

高容量储氢材料在燃料电池和储热等方面有着良好的潜在应用.从高体积密度(kg/m3)和高储氢质量分数两个方面综述了高容量储氢材料的国内外研究近况.从材料组成、制备工艺、材料的组织结构以及催化剂应用等方面重点评述了Mg2FeH6、LiBH4、NaBH4、LiAlH4、NaAlH4等储氢材料的研究进展,指出高容量储氢材料今后中长期研究的重点是NaAlH4、Mg2 FeH6等络合氢化物以及催化剂.