(JAERI) The Third International Symposium on Advanced Nuclear Energy Research “Global Environment and Nuclear Energy”

Nodular corrosion on the surface of Zircaloy-2 fuel cladding tubes was investigated and the effects of irradiation time, linear heat rate, and fast neutron flux on the nucleation and growth of nodules were statistically studied. The nucleation and growth rate of nodules were found to decrease with irradiation time. Then, the nodular corrosion resistance of neutron irradiated Zircaloy-2 was examined in an out-of-pile corrosion test, and neutron irradiation was seemed to improve the Zircaloy-2 nodular corrosion resistance. Finally, microstructural evolution of Zircaloy-2 during neutron irradiation was investigated and radiation-induced dissolution of inter-metallic precipitates was confirmed. It was proposed that an increase of the dissolved alloying element concentration should improve the nodular corrosion resistance of Zircaloy-2 and re- strain the nucleation and growth of nodules on Zircaloy-2 fuel cladding tubes during irradiation in BWRs.     

Residual stresses and tetragonal phase fraction characterisation of corrosion tested Zircaloy-4 using energy dispersive synchrotron X-ray diffraction

To improve the understanding of the oxidation mechanism in zirconium alloys for fuel clad applications, detailed residual stress and phase fraction analysis was carried out for the oxides formed on Zircaloy-4 after autoclave exposure at 360 °C for various times by means of synchrotron X-ray diffraction. In a post-transition sample (220 days), significant stress variation through the oxide thickness was found for the monoclinic phase in individual oxide layers, with maximum in-plane compressive stresses located towards the metal–oxide interface and a discontinuity in the residual stress profile. The depth of this discontinuity matched well with the depth at which electron microscopy analysis showed an interface between two distinct oxide layers. Analysis of the tetragonal phase with exposure time demonstrated changes of the total volume of tetragonal phase before and during transition. These observations are put into the context of residual stress evolution presented previously, to provide further insight into the importance of phase transformations and residual stresses in determining the corrosion kinetics of Zr alloys.     

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.     

Mössbauer spectroscopy of tin in unirradiated and neutron irradiated Zircaloys

Mössbauer spectroscopy of the 23.9 keV γ-rays in ¹¹⁹Sn nuclei was applied to study Zircaloy-2, Zircaloy-4, and other tin-bearing zirconium-based alloys of interest to nuclear power technology. Zircaloys are extensively used in nuclear reactors as fuel cladding. In CANDU reactors, Zircaloys are also used as major structural components such as calandria tubes, and were used until the late 1970's as pressure tubes (now replaced by Zr–2.5Nb alloy). Unirradiated specimens of these alloys, as well as radioactive specimens, both neutron-irradiated in high-flux test reactors and extracted from nuclear power-reactor components after many years of service, were examined. The obtained spectra consistently showed tin in substitutional solid solution in α-Zr, whereas no evidence was found of metallic Sn or intermetallic Zr₄Sn precipitates. In oxide scrapes removed from Zircaloy-2 pressure tube of one of CANDU reactors, where the alloy was exposed for about 10 years to pressurized heavy water coolant at temperatures of ∼280°C, a considerable fraction of tin was found in the Sn(IV) state, in the form that coincides with the state of tin in stannic oxide, SnO₂. The same form of tin was identified in filterable deposits in the primary heavy water coolant of CANDU reactors. For comparison, in Zircaloy heated in air, SnO₂ was formed only at temperatures above 500°C.     

Crystal structure and grain size of Zr oxide characterized by synchrotron radiation microdiffraction

Zirconium oxides formed on Zirclaoy-4 and Zr-1.5Nb (in wt%) were characterized by the microbeam X-ray diffraction using a synchrotron radiation. The phase fraction and the grain size were determined as a function of the position in the oxide. It was found that Zr-1.5Nb showed the better corrosion resistance than Zircaloy-4 in 360 °C pure water although the tetragonal phase was more stabilized to a further distance from the metal/oxide interface in the oxide of Zircaloy-4. The calculation of the grain size revealed that the oxide of the Zr-1.5Nb had larger grains than that of Zircaloy-4 with the tetragonal phase being smaller than the monoclinic one. It seems reasonable to suppose that the superior corrosion resistance of Zr-1.5Nb was attributed to the lager grain size of the oxide in which the oxygen diffusion is expected to be lowered when compared to the smaller grain size of the oxide on Zircaloy-4.     

Microstructural analysis and XPS investigation of nodular oxides formed on Zircaloy-4

In order to study nodular oxidation behavior under LOCA conditions, Zircaloy-4 cladding tubes were oxidized in high-temperature steam ambience and the growth evolution and microstructural properties of the oxides formed on the Zircaloy-4 surfaces were then investigated. Optical metallography showed different growth behaviors of uniform black oxide and localized whitish nodular oxide. The changes in composition and chemical states of the oxides formed on the Zircaloy-4 surface were investigated using X-ray photoelectron spectroscopy (XPS). Anion-deficient and non-stoichiometric ZrO was the main species in initial black uniform oxide surface, while stoichiometric ZrO₂ was the main oxide species in the whitish nodular oxide. A stoichiometric composition in the nodular oxide resulted in a decrease in plasticity of the oxide layers. Unlike black uniform oxide, XPS spectrum from the nodular oxide showed clear Sn photopeak, which indicates that Sn species were observed in the nodular oxide only. As a result, it is concluded that the decreased plasticity and localized Sn additives may be the causes of nodular oxide initiation under LOCA conditions.     

加工工艺对Zr—4管抗疖状腐蚀的影响

应用高压釜，金相及电子显微镜等研究了不同加工工艺的Ｚｒ－４包壳的疖状腐蚀行为。结果表明，改进工艺加工的管材比常规工艺加工的管材有更优良的抗疖状腐蚀性能，去应力试样比再结晶试样有较强的抗疖状腐蚀能力，影响Ｚｒ－４合金抗腐蚀性能的主要因素是Ｆｅ和Ｃｒ合金元素在α－Ｚｒ中的固溶含量，而不是第二相粒子的大小。     

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.     

Changes in chemical structure of oxidation reaction layers of Zircaloy-4 and Ti by micro X-ray diffractometry

A micro X-ray diffractometer with a micrometer sized beam concentrator was developed to investigate the changes in the chemical structures of oxide layers for Zr-based alloys (Zircaloy-4) and Ti metal from the center of the cross section to the surface. Zircaloy-4 and Ti metal were chosen because of their use as a fuel cladding and a heat exchange tubing in a nuclear reactor, respectively. The diffraction patterns were obtained from the cross sectional specimens of the oxidized Zircaloy-4 and Ti metal at 50 μm intervals. For the cross section of Zircaloy-4, Zr metal (hexagonal) was identified in the center, ZrO_2−x (hexagonal, about 200 μm in thickness) inside the edge and ZrO₂ (monoclinic, about 400 μm in thickness) at the edge. In the case of Ti metal, Ti metal (hexagonal) was identified in the center, TiO (cubic, about 200 μm in thickness) inside the edge and rutile-TiO₂ (tetragonal, about 230 μm in thickness) at the edge. From this study, it was concluded that the intermediate phase formed between the fuel and the cladding can be identified by the micro-XRD system.     

Zr-Sn-Nb-Fe-V合金包壳管加工过程中第二相的演变

采用扫描电子显微镜（SEM）和透射电子显微镜（TEM）研究了Zr-0.2Sn-1.3Nb-0.2Fe-0.05V合金经热挤压、冷轧、中间退火包壳管坯以及经终轧及最终退火后成品管材第二相特征。结果表明,热挤压产生的β-Zr及第二相沿管坯轴向呈流线状分布,随着冷轧和退火的进行,亚稳相β-Zr发生分解,第二相逐渐均匀化,最终呈细小、均匀、弥散分布。合金成品管材第二相主要为BCC结构的β-Nb,含有少量FCC结构的Zr(NbFeV)₂。加工过程中析出相的平均直径变化不大,均小于100 nm。合金包壳管第二相尺寸分布与热处理过程中含Nb第二相溶解析出直接相关。     

Corrosion behavior of Zr alloys with a high Nb content

The corrosion behavior of the Zr alloy with a high Nb content was evaluated in the water loop system containing 2.2 wppm Li and 650 wppm B. The characteristics of the precipitates were analyzed by transmission electron microscopy (TEM) and the oxide was characterized by an X-ray diffraction method using a synchrotron radiation source. On the basis of the results obtained by these measurements, the relationship among the oxidation behavior, the precipitate characteristics and the oxide properties was discussed. It was shown that the Cu addition was of benefit to the corrosion resistance of the Zr alloy with a high Nb content and the corrosion resistance of the Cu-containing alloy (Zr–1.5Nb–0.5Sn–0.2Fe–0.1Cu) was superior to that of the Cr-containing alloy (Zr–1.5Nb–0.5Sn–0.2Fe–0.1Cr). The fine β-Nb precipitates were found more frequently in the Cu-containing alloy than the Cr-containing alloy when heat-treated in the same condition. The fraction of the tetragonal zirconia in the region of the metal/oxide interface was higher in the Cu-containing alloy than the Cr-containing alloy, suggesting that the stabilization of the tetragonal phase in the oxide was promoted more when the smaller precipitates are incorporated into the oxide. It is concluded that the fine distribution of β-Nb is desirable for stabilizing the tetragonal phase in the oxide, thereby increasing the corrosion resistance of the Zr alloy with a high Nb content.     

N36锆合金包壳堆内腐蚀模型研究

本文利用池边检查数据,基于阿累尼乌斯方程建立了N36锆合金包壳堆内腐蚀最佳估算模型。由于缺乏腐蚀转折前数据,N36锆合金包壳腐蚀转折前氧化膜厚度只是时间的函数,腐蚀转折后氧化膜厚度是包壳温度和时间的函数。通过在最佳估算腐蚀模型上添加工程因子,建立了不同加工工艺N36锆合金包壳腐蚀模型。N36锆合金包壳腐蚀包络模型在最小腐蚀转折点的基础上建立。模型验证结果表明,N36锆合金包壳腐蚀模型与验证数据符合较好,能够用于N36锆合金堆内腐蚀行为模拟。     

Irradiation Effects on Corrosion Resistance and Microstructure of Zircaloy-4

In order to investigate irradiation effects on nodular corrosion resistance of Zircaloy-4, an out-of-pile corrosion test was conducted using Zircaloy-4 specimens cut from the channel box of a fuel assembly irradiated in the BWR (Monticello reactor) up to the neutron fluence of 1.53×10²⁶ n/m² (E>1MeV). The corrosion test was carried out in high pressure steam of 10.3 MPa at 783 K for 24 h. No nodules appeared on the specimens cut from welded areas of the channel box and nodular corrosion resistance tended to be better with increasing neutron fluence. Microstructural evolution in the form of irradiation-induced release of Fe atoms from Zr(Fe, Cr)2 type precipitates was detected by an analytical electron microscope. It was found that the higher the concentration of dissolved Fe and Cr in the grains of Zircaloy-4, the better the nodular corrosion resistance.     

Structure of zirconium alloy oxides formed in pure water studied with synchrotron radiation and optical microscopy: relation to corrosion rate

A detailed study was undertaken of oxides formed in 360 °C water on four Zr-based alloys (Zircaloy-4, ZIRLO™,¹ Zr-2.5%Nb and Zr-2.5%Nb-0.5%Cu) in an effort to relate oxide structure to corrosion performance. Micro-beam X-ray diffraction was used along with transmitted light optical microscopy to obtain information about the structure of these oxides as a function of distance from the oxide-metal interface. Optical microscopy revealed a layered oxide structure in which the average layer thickness was inversely proportional to the post-transition corrosion rate. The detailed diffraction studies showed an oxide that contained both tetragonal and monoclinic ZrO₂, with a higher fraction of tetragonal oxide near the oxide-metal interface, in a region roughly corresponding to one oxide layer. Evidence was seen also of a cyclic variation of the tetragonal and monoclinic oxide across the oxide thickness with a period of the layer thickness. The results also indicate that the final grain size of the tetragonal phase is smaller than that of the monoclinic phase and the monoclinic grain size is smaller in Zircaloy-4 and ZIRLO than in the other two alloys. These results are discussed in terms of a model of oxide growth based on the periodic breakdown and reconstitution of a protective layer.     

M5合金的堆内外性能概述

概述了法国法马通公司开发的新型燃料包壳材料 M5合金在堆内外的腐蚀、吸氢、显微组织、蠕变、辐照生长等性能。从已获得的堆内数据证明， M5合金包壳在抗腐蚀、吸氢、蠕变、辐照生长方面大大优于最佳化的 Zr- 4合金包壳。可以预计， M5新合金包壳能满足燃耗达 65GW· d· t－ 1的设计要求。在法马通推出的 PWR燃料组件 AFA- 3G已采用了 M5合金包壳。     

Zr-4合金a-Zr固溶体中的Fe、Cr含量分析

将电化学分离Zr-4合金中Zr(Fe,Cr)2第二相粒子的技术和原子吸收光谱分析技术相结合,建立了一种分析Zr-4合金a-Zr固溶体中Fe、Cr含量的新方法,并用这种新方法分析了不同热处理状态下Zr-4合金a-Zr固溶体中的Fe、Cr含量。分析结果表明,随着淬火温度的增加,a-Zr固溶体中的Fe、Cr含量和Fe/Cr比值均增加,而Zr(Fe,Cr)2第二相粒子中的Fe/Cr比值则相应降低。结合以前的工作,可得结论:过饱和固溶在a-Zr固溶体中的Fe、Cr含量对Zr-4合金的耐水侧腐蚀性能有重要影响。     

The effect of alloying modifications on hydrogen uptake of zirconium-alloy welding specimens during corrosion tests

The hydrogen uptake behavior during corrosion tests for electron beam welding specimens made out of Zircaloy-4 and zirconium alloys with different compositions was investigated. Results showed that the hydrogen uptake in the specimens after corrosion tests increased with increasing Cr content in the molten zone. This indicated that Cr element significantly affected the hydrogen uptake behavior. Fe and Cr have a low solubility in α-Zr and exist mainly in the form of Zr(Fe,Cr)₂ precipitates, which is extremely reactive with hydrogen in its metallic state. It is concluded that the presence of Zr(Fe,Cr)₂ second phase particles (SPPs) is responsible for the increase in the amount of hydrogen uptake in the molten zone of the welding samples after corrosion, as Zr(Fe,Cr)₂ SPPs embedded in α-Zr matrix and exposed at the metal/oxide interface could act as a preferred path for hydrogen uptake.     

Examination of the chemical composition of irradiated zirconium based fuel claddings at the metal/oxide interface by TEM

Detailed post-irradiation examinations have been performed at PSI on three fuel rods with differing cladding materials revealing different corrosion behaviour. The rods had been irradiated for 3-5 cycles at Gösgen nuclear power plant (pressurised water reactor), Switzerland. As zirconium corrosion is proceeding at the metal/oxide interface, extended micro-structural analyses were performed by transmission electron microscopy (TEM), expecting to possibly reveal phenomena explaining the varying corrosion resistance. This paper reports on the distribution of oxygen at the metal/oxide interface examined by energy dispersive X-ray spectroscopy (EDS) in TEM, while other micro-structural investigations have been published earlier [1]. In order to get some statistical confidence in the analyses, three neighbouring TEM samples of each cladding variant were studied. The oxygen concentration profiles of the three alloys (i.e. low-tin Zircaloy-4, Zr2.5%Nb and extra low-tin (Sn 0.56%)) both in the oxide and metal close to the metal/oxide interface are compared. The results of the examinations show the composition of the oxide in the vicinity of the interface to be sub-stoichiometric for all three materials, indicating an oxide layer adjacent to the interface, with diffusion-controlled access of oxygen to the metal/oxide interface. The metallic parts show highest oxygen concentrations at the metal/oxide interface which are reduced towards the bulk metal, pointing towards the expected second diffusion-controlled process leading to α-Zr (O). Based on the experimental results values for the diffusion coefficients in the range of 0.8-6.0 × 10⁻²⁰ m² s⁻¹ are estimated for the oxygen dissolution process, the diffusion coefficient in Zircaloy-4 being six times higher than for the other two less corroding alloys. This finding is in contradiction with the present assumptions about the corrosion mechanism, and confirms the expected but not so far reported diffusion controlled oxidation of different zirconium alloys. It also points towards a corrosion rate that is at least partly governed by the diffusion coefficient of oxygen in metal that is different for different alloys, unlike what has been assumed till present.     

Fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

ABSTRACT

To evaluate the fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident (LOCA) conditions, laboratory-scale integral thermal shock tests were performed using the following advanced fuel cladding tubes with burnups of 73–85 GWd/t: M-MDA^TM, low-tin ZIRLO^TM, M5®, and Zircaloy-2 (LK3). In total eight integral thermal shock tests were performed for these fuel cladding tube specimens, simulating LOCA conditions including ballooning and rupture, oxidation, hydriding, and quenching. During the tests, the specimens were oxidized to 10%–30% equivalent cladding reacted (ECR) at approximately 1473 K and were quenched under axial restraint load of approximately 520–530 N. The effects of burnup extension and use of the advanced fuel cladding tubes on the ballooning and rupture, oxidation, and hydriding under LOCA conditions were inconsiderable. Further, the high-burnup advanced fuel cladding tube specimens did not fracture in the ECR values equal to or lower than the fracture limits of the unirradiated Zircaloy-4 cladding tube reported in previous studies. Therefore, it can be concluded that the fracture limit of fuel cladding tubes is not significantly reduced by extending the burnup to approximately 85 GWd/t and using the advanced fuel cladding tubes, though it slightly decreases with increasing initial hydrogen concentration.     

Oxidation properties of Zr–Nb alloys at 973–1273 K in air

In order to investigate the oxidation behavior of LWR cladding materials under the condition of reactor accidents, e.g. LOCA, Zr–Nb alloys with 1–10 wt%Nb and Zircaloy-4 (0 wt%Nb) were oxidized at 973–1273 K in dry air. The weight gain due to oxidation increased with Nb content at 973 and 1073 K was the smallest for 2.5 wt%Nb at 1173 and 1273 K. The oxidation kinetics obeyed the parabolic rate law without a few cases, e.g. 6–10 wt%Nb and 1273 K. The parabolic rate constant at high temperatures had the somewhat low activation energy compared to that at low temperatures. These results implied that such oxidation behaviors of Zr–Nb alloys related to the lattice structures of oxide films as well as underlying metal during oxidation. Especially at high temperatures, 6ZrO₂–Nb₂O₅ compound might promote the oxidation of Zr–Nb alloys with high content of Nb.