Cladding Tube Research Articles

Influence of cladding tube structure on transmission characteristics of negative curvature terahertz fiber

A novel type of polarization negative curvature terahertz fiber has been designed, which has a cladding region composed of two long semi-circular cladding tubes in the horizontal direction and four long semi-elliptical cladding tubes in the oblique direction. The optimal fiber dimensions are determined to investigate the internal relationship between polarization characteristics and fiber cladding structure. The transmission performances of the negative curvature fiber are investigated by varying the thickness of fiber cladding tubes and adjusting the structural characteristics of cladding tubes in different directions. Relatively good transmission performances can be obtained with a cladding tube thickness of 80 μm. The birefringence can be stably maintained at the order of 10−4, and the confinement loss of the x-polarized mode at 2.55 THz is 0.036 dB cm−1. Moreover, the dispersion coefficient of the two polarized modes is stable between ±0.6 ps (THz·cm)−1 in the 2.3–2.6 THz band. The bent performance of the designed fiber at 2.52 THz is also discussed. The results show that the fiber has low bending sensitivity, and can still maintain excellent birefringence and low confinement loss in the bending state. For the y polarization, the birefringence can be basically stabilized at the order of 10−4 within a bending radius of 10–60 cm.

Analysis of multi-scale failure behavior of SiCf/SiC composite clad tube under flexible clamping damage

The intricate spatial fiber winding structure of SiCf/SiC ceramic matrix composite (CMC) cladding tubes makes it difficult to investigate in depth the inherent evolutionary mechanism of damage failure under specific operating conditions. Therefore, in this study, based on X-ray computed tomography (X-CT) reconstruction and cross scale numerical simulation methods, combined with clamping failure experiments under specific conditions for the cladding tubes, we conduct an in-depth analysis of the material's damage, failure, and mechanical behavior in different clamped regions. The research indicates that variations in the clamped regions have a minimal impact on the overall mechanical feedback of the material, and the maximum clamping load remains essentially around 500 N. The tube damage exhibits a unique " regional progressive damage evolution mechanism," where stress differentials of 108.49–115.33% appear between the inner and outer surfaces in different regions, leading to variations in the initiation and propagation of cracks within these areas. Furthermore, based on the minimum potential energy principle and semi-geodesic line theory, the intricate macroscopic tube structure formed by the spatial stacking and winding of SiCf/SiC cladding tube fibers is accurately reconstructed, which is combined with macroscale and microscale local feature models to capture the fibre/matrix morphology of the material with dynamic mechanical feedbacks, and experimentally validated in this work. The results indicate that the numerical model can effectively characterize the complex mechanical damage morphology and progressive failure lows of SiCf/SiC CMC at multiscale. Simultaneously, it is observed that stress concentrations in the fiber crossover region result in a load-bearing capacity 112–283% higher than in other areas. This phenomenon often leads to the first occurrence of failure in this region after clamping overload, exhibiting more pronounced damage characteristics.

Fretting wear behaviors of Cr-coated Zr–1Nb alloy cladding tube mated with Zr-4 alloy dimple under different tangential displacements

The fretting wear behaviors and wear mechanism of the Cr-coated Zr–1Nb alloy cladding tube mated with Zr-4 alloy dimple were investigated under different tangential displacements. With the increasing tangential displacement, the fretting running status changes from mixed fretting regime to gross slip regime, the friction coefficient increases first and then decreases, and the removed material volume, maximum wear depth, and wear coefficient of the Cr-coated Zr–1Nb alloy cladding tube increase. Fouvry's criteria were used to verify the correctness of the fretting running status. The primary wear mechanisms in the mixed fretting regime are adhesive wear and oxidative wear, while those in the gross slip regime are delamination wear and oxidative wear. The wear mechanism of the dimple is mainly delamination wear. Cr coating inhibits the wear of the cladding tube. After the coating is completely removed, the volume of material removed from the cladding tube increases dramatically, eventually exceeding the volume of material removed by dimple.

Prediction of ‘effective’ corrosion rates by in-situ internal pressure creep testing in LBE for Ti-stabilised DIN 1.4970 (15-15Ti) austenitic stainless steel cladding tubes

ABSTRACT In this paper, a new high-temperature methodology for determining the ‘effective’ corrosion rates of internal pressure-tested cladding tubes, based on the reduced creep failure times under exposure, is presented. The methodology has been developed to estimate the combined damage accumulation of corrosion and creep and to establish temperature limits for the use of 1515Ti claddings in high-temperature Lead-Bismuth-Eutectic (LBE). The results are compared to static corrosion rates determined from earlier test campaigns in stagnant LBE. All tests are conducted on 24% cold worked 1515Ti cladding tubes that are the SCK CEN reference material for the MYRRHA design efforts. The virgin material creep model that is the basis for the corrosion rate estimation, has been fitted on an extended creep failure database and a life-fraction rule is applied to accommodate for the life reduction caused by the sought ‘constant’ effective corrosion rate.

Dislocation Hardening in a New Manufacturing Route of Ferritic Oxide Dispersion-Strengthened Fe-14Cr Cladding Tube.

The microstructure evolution associated with the cold forming sequence of an Fe-14Cr-1W-0.3Ti-0.3Y2O3 grade ferritic stainless steel strengthened by dispersion of nano oxides (ODS) was investigated. The material, initially hot extruded at 1100 °C and then shaped into cladding tube geometry via HPTR cold pilgering, shows a high microstructure stability that affects stress release heat treatment efficiency. Each step of the process was analyzed to better understand the microstructure stability of the material. Despite high levels of stored energy, heat treatments, up to 1350 °C, do not allow for recrystallization of the material. The Vickers hardness shows significant variations along the manufacturing steps. Thanks to a combination of EBSD and X-ray diffraction measurements, this study gives a new insight into the contribution of statistically stored dislocation (SSD) recovery on the hardness evolution during an ODS steel cold forming sequence. SSD density, close to 4.1015 m-2 after cold rolling, drops by only an order of magnitude during heat treatment, while geometrically necessary dislocation (GND) density, close to 1.1015 m-2, remains stable. Hardness decrease during heat treatments appears to be controlled only by the evolution of SSD.

Effect of Sputtering Bias Voltage on the Microstructure and Lead-Bismuth Eutectic Corrosion Resistance of the FeCrAlY Coating

FeCrAlY coatings deposited on the outer surface of the cladding tube by the magnetron sputtering with varying bias voltages (0 V to −500 V) were corroded at 550°C to 600°C in oxygen-saturated and oxygen-poor lead-bismuth eutectic for 500 h to 1,000 h. The increased bias voltage led to the Al content decreased and the Y content increased. The coating changed from columnar crystals to equiaxed crystals with increased bias voltage. The decreased bias voltage makes an improved corrosion resistance. The fundamental process governing the evolution of microstructure and corrosion resistance in the FeCrAlY coatings was discussed.

A study on the fracture pattern change of high-burnup fuel cladding failed by pellet-cladding mechanical interaction failure under reactivity-initiated accident conditions

ABSTRACT In a part of the historical reactivity-initiated accident (RIA) simulated tests on high-burnup fuels performed at the Nuclear Safety Research Reactor, the fuel failure caused by pellet-cladding mechanical interaction (PCMI) led to splitting into upper- and lower-part pieces or even fragmentation of the cladding tube. A massive release of fuel fragments accompanied this fracture pattern change from previously known axial cracks and thus identified as a potential concern in safety evaluation regarding core coolability. Dedicated out-of-pile mechanical tests were performed with unirradiated Zircaloy-4 cladding specimens to clarify the condition of such fracture pattern change. The specimens were pre-hydrided and subjected to loading with axial-to-hoop strain ratios of ℇz/ℇθ = 0.5–1.25, simulating the effects of hydrogen embrittlement and pellet-cladding mechanical bonding of high-burnup fuels, respectively. The results indicate that higher biaxiality of the loading and lower ductility (failure strain level) assist the fracture pattern change. This study proposes a conservative criterion that a PCMI failure splits the cladding tube into more than two pieces when strain ratio ℇz/ℇθ >0.75 and a concurrent hoop strain < 1.7% at the failure instant.

Reaction Capsule Design for Interaction of Heavy Liquid Metal Coolant, Fuel Cladding, and Simulated JOG Phase at Accident Conditions

High temperature corrosion of fuel cladding material (15-15Ti) in high burn-up situations has been an important topic for molten metal-cooled Gen-IV reactors. The present study aims to investigate the simultaneous impact of liquid lead (coolant side) and cesium molybdate (fuel side) on the cladding tube material. A capsule was designed and built for experiments between 600 °C and 1000 °C. In order to simulate a cladding breach scenario, a notch design on the cladding tube was investigated pre- and postexposure. Material thinning by corrosion and leaching at temperatures ≥ 900 °C caused breaches at the notches after 168 h exposure. The temperature dependent cladding thinning phenomenon was used for kinetic interpretation. As the first of a two-part study, this paper will focus on the exposure capsule performance, including metallographic cross-section preparation and preliminary results on the interface chemistry.

Formation mechanism, microstructure and mechanical properties in current-introduced solid-state welding of dissimilar zirconium alloy

The cladding tube (Zr-Nb alloy) and the end plug (Zr-Sn alloy) were successfully joined using current-introduced solid-state welding. During the welding process, the metals undergo severe plastic deformation and were joined by dynamic recrystallization. With the increase of the forging force, the blasting force of the joint changes little. The joint exhibited a blasting force ranging from 8086 to 9037 N, and the fracture occurred in the cladding tube, away from the weld zone. The blasting fracture mode of joint is ductile fracture.

The Precipitation and Redistribution of Alloying Element in Zircaloy-4 Cladding Tube Oxidized in High-Temperature Steam

The Precipitation and Redistribution of Alloying Element in Zircaloy-4 Cladding Tube Oxidized in High-Temperature Steam

Small punch test for investigating circumferential creep in cladding tubes

Several methods have been developed to determine the axial creep properties of nuclear fuel cladding tubes. However, the anisotropic creep properties of nuclear fuel cladding tubes requires that their circumferential creep behaviour be evaluated as well. Nonetheless, this task remains extremely challenging owing to need to employ unaltered cylindrical cladding tube specimens directly and the general low accuracy of the methods developed for this purpose. The present work addresses these issues by developing a modified small punch creep test (SPCT). Equations are proposed to calculate the creep stress and strain of the modified SPCT. The accuracy of the equations is analysed using the finite element method. Finally, the methodology is applied to investigate the circumferential creep behaviour of an FeCrAl nuclear fuel cladding tube experimentally. These results clearly demonstrate that the modified SPCT is suitable for investigating the creep behaviour of irregularly shaped material specimens with anisotropic creep properties. Moreover, the semi-cylindrical specimens employed in the modified SPCT also enable the effective fracture strain of the specimen to be easily determined.

Determination of critical material damage parameter for Indian clad tube burst behavior under severe accident scenario

During loss of coolant accident scenario in PHWRs, heat removal from clad tubes containing the uranium oxides fuel may get severely impaired. This results in increase in temperature of the fuel clad and subsequent increase in the clad tube internal pressure. This causes the clad tube to deform and balloon leading to the increase in diameter of the fuel clad. After a critical value of deformation is reached, the clad tube gets burst. The burst process is due to initiation of crack in the clad due to creep deformation process in the presence of internal pressure and high temperature environment. In this work, experimental data of fuel clad bursts from literature have been used to evaluate a critical damage parameter that can be used to predict the onset of burst phenomenon of clad tubes. Finite element modelling of the fuel clad has been carried out to simulate the ballooning deformation of the fuel clad tubes due to creep. Norton’s creep model has been used in the simulation. The critical damage parameter evaluated, using analysis and experimental data, has been used subsequently to predict burst time of Zircaloy-4 fuel clad when subjected to different types of loading environments. An exponential decay type correlation has been obtained from the finite element analysis results which can be used by safety analysts to predict burst pressure of Zircaloy-4 fuel clad of Indian PHWR as a function of burst temperature or vice-versa.

Deformation law and bonding mechanism of 45 carbon steel/316L stainless steel cladding tubes fabricated by three-roll skew rolling bonding process

Realizing the continuous forming and metallurgical bonding in bimetallic cladding tubes with high-strength and corrosion-resistance is the key bottleneck to be solved urgently. In this study, 45 carbon steel/316L stainless steel (CS/SS) cladding tubes were fabricated by the three-roll skew rolling bonding process under different radial reduction rates. The deformation law, interfacial microstructure characteristics, and deformation mechanism were analyzed. With the increase of the radial reduction rates from 11.7 % to 19.3 %, the 45 CS/316L SS cladding tubes were elongated along with the diameter and wall thickness decrease. Besides, the wall thickness ratio between 45 CS and 316L SS was unchanged due to the coordinated deformation. Tensile shear fracture locations of CS/SS cladding tubes were on the 45 CS matrix, and the average diffusion distances in the longitudinal section and the cross section were consistent, proving that uniform metallurgical bonding was achieved. Besides, average decarburized layer thickness decreased and average diffusion distances increased with the increase of radial reduction rates. Furthermore, spherical and fibrous Mn-Si-Cr-O oxides, generated in the insulation heating process, were dispersed near the bonding interface. The refined grains were formed close to the interface, and deformation twins were formed away from the interface on the 316L SS side, whereas only elongated ferrite grains evolved due to the decarburization adjacent to the interface on the 45 CS side. Therefore, the interfacial bonding mechanism was illustrated as three stages, namely the insulation heating stage, the skew rolling bonding stage, and the air cooling stage. The three-roll skew rolling bonding process provides a high-efficiency method to prepare CS/SS cladding tubes with uniform metallurgical bonding, which adopts axial forming and contributes to breaking through the product length size limit.

High Temperature Corrosion Behavior of 15-15Ti Cladding Tube Material in Contact with Liquid Lead, Outside, and Cs2MoO4, Inside

Lead-Cooled Fast Reactors are one of the emerging new technologies connected to the Generation IV reactor designs. Even though the reactor features are extensive and beneficial, some technical and regulatory difficulties remain that hinder the deployment. One of them is the corrosion behavior of fuel claddings in high burn-up and elevated temperatures, as fuel claddings are a part of accident barriers. In this study, 15-15Ti austenitic steel was the investigated cladding material.As this cladding would be used in a tube form, corrosion attack from both sides, inside and outside is studied in contact to their respective environments. From the outside of the tube, the proposed coolant liquid lead is the corrosive substance. At the inside of the tube, the major corrosive species in our test setup is, for simplicity reasons, Cs2MoO4. However, this choice has been made from some considerations regarding the fuel evolution upon burn-up.In a high burn-up situation, the generated fission products would start to accumulate to the gap between the fuel pellet and the cladding tube. These accumulated products are the so-called “Joint-Oxide Gain” (JOG) phases, and they are the corrosive substances in question when discussing the corrosion attack from the inside of a cladding tube. Cs2MoO4 has been identified as one of the main components of JOG phases and thus will be focused on in this study.A new capsule was designed to investigate the dual-atmosphere corrosion of cladding tubes. Unexposed capsule parts (a) and the basic schematic of the assembled capsule (b) can be seen in the Figure below. The cladding tube material itself is used as the container for JOG as it is filled with Cs2MoO4 and sealed on both ends. This assembly is then placed inside a bigger capsule, which is filled with lead powder to submerge the cladding tube assembly. All assemblies were done in Ar-filled glovebox. The capsules were then exposed to 600-1000 oC for 52 to 168 hours, which was followed by cutting and epoxy embedding to investigate the cross-sections. Figure 1

Structure and Oxidation Behavior of a Chromium Coating on Zr Alloy Cladding Tubes Deposited by High-Speed Laser Cladding

Structure and Oxidation Behavior of a Chromium Coating on Zr Alloy Cladding Tubes Deposited by High-Speed Laser Cladding

Preparation and properties of water-soluble coatings resistant to scratching on the surface of nuclear cladding

As the second barrier of radioactivity in the reactor core, Zr alloy nuclear cladding tubes are easily to be scratched by the fixed grid during assembly process. In order to protect them from possible damage during their insertion through the skeleton, it is proposed to use water-soluble coatings with excellent scratch resistance to protect nuclear cladding tubes, and then wash away the coating to prevent the impeding the normal operation of the assembly. In this study, polyacrylate emulsion, neutralizer and filler were used to synthesize the coating. The chemical composition, thermal stability and hydrophilic performance of the coating were analysed by Fourier Transform Infrared Spectroscopy (FT-IR), thermal gravimetric analysis (TGA) and water contact angle (WCA) respectively. TGA analysis showed that the weight loss rate of the coating was less than 10 % at 200 °C, indicating the good thermal stability of coating at 200 °C. The water contact Angle of the coating was reduced to 32.78°, indicating that the coating had good hydrophilicity. Scanning electron microscopy displayed that the micro-structure of coating was dense, which was one reason of the good scratching resistant. By adjusting the curing conditions, the optimized preparation parameters of the coating were obtained. When the curing temperature was 90 °C and the curing time was 30 min, the coating had more optimized scratch resistance and water solubility.

High frequency acoustic microscopy imaging of pellet cladding interface in nuclear fuel rods

In Pressurized Water Reactors (PWRs), fuel rods play a crucial role in generating nuclear energy. These rods consist of ceramic pellets, such as UO2 or (U,Pu)O2, enclosed in a zircaloy cladding tube, leaving an initial gap between the pellets and the cladding. As the reactor operates and the fuel undergoes irradiation, both the ceramic pellets and the zircaloy cladding experience transformations, causing the gap between them to gradually close. This phenomenon has a significant impact on the thermomechanical behavior of the fuel rod.Understanding the nature of the bonding that occurs during irradiation is essential for ensuring the safe and efficient operation of the reactor. To investigate the evolution of the contact state between the fuel pellets and the cladding during irradiation, a detailed analysis of the pellet-cladding interface after irradiation is necessary. However, traditional examination methods might be destructive or incapable of providing the desired level of precision and resolution.The Institute of Electronic and Systems at the University of Montpellier (IES – UMR CNRS 5214), in collaboration with the Alternative Energies and Atomic Energy Commission (CEA) and Electricité de France (EDF), has developed a specialized high-frequency acoustic microscope for imaging and non-destructively inspecting the pellet/cladding interface. The design of the acoustic microscope takes into account the complexity of the fuel rod's structure and the challenges associated with imaging the pellet/cladding interface by utilizing high-frequency ultrasound.In this paper, we present the ability of this acoustic microscope to acquire 2D images with controlled displacements of the sample rod along both its axial and circumferential directions thanks to a card with a high sampling frequency reaching 2 GHz. This capability is crucial because the geometrical, chemical, and mechanical properties of the fuel pellet-cladding contact are not uniform in these directions. By obtaining detailed acoustic images, we can identify specific regions where the fuel pellets and the cladding were in contact during irradiation. In this research, a resolution study is carried out to validate the microscope's ability to investigate the fuel rod and achieve the desired resolutions.Testing on real samples requires a specific configuration of the microscope, which must be adapted to the irradiation conditions. This is why, before proceeding to this stage, it is necessary to carry out tests on representative samples to validate the achievement of the desired resolution. So we’re also presenting the first acoustic images obtained on the zircaloy alloy claddings.

Accelerated Formation of Oxide Layers on Zircaloy-4 Utilizing Air Oxidation and Comparison with Water-Corroded Oxide Layers.

For the dry storage of Canada Deuterium Uranium (CANDU) spent nuclear fuels, the integrity of Zircaloy-4 fuel cladding has to be verified. However, the formation of ~10 µm-thick oxide layers in typical CANDU reactor operating conditions takes several years, which makes sample preparation a slow process. To overcome such limitations, in this study, an accelerated formation of an oxide layer on Zircaloy-4 cladding tube was developed with a combination of high-temperature water corrosion (HT-WC) and air oxidation (AO). First, Zircaloy-4 tubes were corroded in oxygenated (2 ppm dissolved oxygen) high-temperature water (360 °C/19.5 MPa) for 500 h. Then, the tubes were air-oxidized at 500 °C for 30 h. Finally, the tubes were corroded again in HT-WC for 500 h to produce ~10 µm-thick oxide layers. The morphology and characteristics of the oxide layer in each step were analyzed using X-ray diffraction, scanning and transmission electron microscopy. The results showed that the oxide layer formed in the accelerated method was comparable to that formed in HT-WC in terms of morphology and oxide phases. Thus, the accelerated oxide formation method can be used to prepare an oxidized Zircaloy-4 cladding tube for CANDU fuel integrity analysis.

Study on frictional behavior of SiCf/SiC composite clad tube clamping condition under nuclear irradiation

Silicon carbide fiber reinforced silicon carbide matrix (SiCf/SiC) composite is the key cladding material of nuclear fuel, which determines the safety and reliability of nuclear fuel storage and transportation. The replacement of its storage and transportation scenario needs to be completed by the manipulator, but the application of SiCf/SiC wear, fracture, and nuclear leakage in the snatching process of brittle-flexible-rigid contact in the irradiation environment has been seriously restricted due to unclear understanding of the damage mechanism. Therefore, the effects of irradiation dose and clamping load on the friction characteristics of the contact interface between SiCf/SiC clad tube are studied in this paper, and the effects of irradiation parameters and clamping force on the static friction coefficient of the contact interface between the clad tube and flexible nitrile are obtained. Based on the Greenwood-Williamson tribological model, a numerical model of the shape and structure of the contact micro-convex at the micro-scale of the clamping interface is constructed by introducing the multi-surface integral, and finally verified by experiments. The research results show that there is a unique “Irradiation suppression zone” under the clamping condition of SiCf/SiC cladding tube under the nuclear irradiation environment, and the growth of static friction coefficient slows down until stagnates after irradiation reaches a certain extent (600 kGy), and there will be a decline when the irradiation dose continues to increase, among which the clamping force of 15.2 N within the irradiation interval of 1,000 kGy can meet the safety of nuclear environment operation. The results of this paper can provide an important theoretical basis and application guidance for the safe operation of SiCf/SiC cladding tubes in the storage and transportation clamping process.

Double nested dual-core negative curvature fiber polarization beam splitter

In this paper, a polarization beam splitter（PBS） based on double nested dual-core negative-curvature fiber (DNHC-NCF) is designed. Two hollow cores are constructed by eight double nested cladding tubes arranged at certain angles. The influences of structural parameters on the performances of the DNHC-NCF PBS have been studied by using the finite element method (FEM). According to the simulation results, the structure of the proposed DNHC-NCF PBS is optimized. The final results indicate that the length of the proposed DNHC-NCF PBS is 1.38 cm and the extinction ratio (ER) is less than −20 dB in the wavelength range of 1.48–1.80 µm (320 nm, covers S + C + L + U bands). Besides, the higher order mode extinction ratio (HOMER) in the wavelength range of 1.3–1.71 µm is greater than 100, showing good single-mode characteristic as well. It is believed that the proposed double nested dual-core negative curvature fiber polarization beam splitter will have important applications in the fields of optical communication system.