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The exploration of flow-induced vibration in fuel rods is of utmost importance as it plays a pivotal role in comprehending and mitigating factors that contribute to fuel failure and reactor shutdown. This understanding is crucial for advancing the nuclear energy industry. To unravel the intricacies of this complex phenomenon, a sophisticated high-fidelity finite element model of fuel rods was meticulously constructed. This model serves as the cornerstone for a computational analysis of flow-induced vibration responses utilizing ANSYS Batch, firmly roots in the principles of random vibration theory. In the pursuit of a profound understanding, support stiffness values to simulate and scrutinize diverse scenarios that fuel rods might encounter were systematically varied. The outcomes of these simulations were meticulously compiled to establish a comprehensive database of flow-induced vibration responses. This database stands as a valuable resource for future research endeavors and engineering applications within the nuclear energy domain. Taking a step forward, an innovative approach was adopted to streamline the analysis process. Leveraging the snapshot matrix derived from the extensive database, a high-fidelity reduced-order model (ROM) was developed. Two data-driven methodologies, namely proper orthogonal decomposition (POD) and dynamic mode decomposition (DMD) methods, were employed to construct the ROM. This ROM enables the rapid reconstruction of flow-induced vibration responses, facilitating efficient analysis and decision-making in real-world applications. A critical facet of this paper involves a comparative analysis of the reconstruction effectiveness of fuel rod flow-induced vibration responses using both the POD and DMD methods. The results of this comparison reveal that, when considering the reconstruction of vibration responses with the same number of modes, the POD method outperforms the DMD method. This finding underscores the importance of selecting appropriate methodologies based on specific objectives and computational efficiency. Furthermore, the results indicate that the DMD method excels not only in efficiently reconstructing the vibration responses of fuel rods but also offers the unique capability to assess the stability of each DMD mode. This dual functionality enhances the overall diagnostic capabilities of the DMD method, providing valuable insights into the dynamic behavior and potential instabilities of the fuel rod system. In conclusion, the exhaustive investigation outlined in this paper not only significantly contributes to the understanding of flow-induced vibration characteristics in fuel rods but also provides a robust framework for developing advanced models and methodologies for future research and practical applications in the nuclear energy industry. The comprehensive nature of our approach ensures that our findings are not only insightful but also applicable in shaping the future of nuclear energy technology.

The technology of nuclear reactors is evolving rapidly, driven by the pursuit of more powerful and efficient systems. The Small Lead-based Reactor (SLR) represents an advanced nuclear reactor design that holds great promise for delivering enhanced power and efficiency. In the context of the SLR's fuel rod, the high-speed coolant flow within the reactor can induce vibration, potentially causing fretting wear and damage to the cladding. This study utilized the Burgreen correlation to establish an equivalence relationship between water and Lead-Bismuth Eutectic (LBE). An experiment was then conducted to simulate axial flow-induced vibration (FIV) of a simply supported fuel rod, employing an equivalent water loop. Flow-induced vibration characteristics in both the time and frequency domains were investigated at different flow speeds. The experimental results revealed a positive correlation between flow velocity and amplitude. The fuel rod exhibited low-frequency vibrations with a random pattern, registering frequencies around 14 Hz. These experimental findings can be leveraged to enhance the accuracy of numerical simulations for axial flow-induced vibration (FIV) of the fuel rod. Subsequently, the revised numerical model was applied to simulate FIV in the LBE environment using computational fluid dynamics (CFD), and the numerical results were found to be in agreement with the experimental data. [ABSTRACT FROM AUTHOR]

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Amidst the global expansion of nuclear power generation, ensuring the integrity of nuclear fuel rods is crucial for the safe operation of nuclear power plants. As a vital component of fuel rods, the detection of defects in the end plugs is a key step in ensuring nuclear safety. Traditional manual detection methods are not only time-consuming and inefficient but also susceptible to subjective influences. To address these issues, this study proposed an automatic detection method for defects in fuel rod end plugs based on deep learning X-ray imaging, aiming to enhance the accuracy and efficiency of detection. The research began by collecting a large number of X-ray images of fuel rod end plugs and preprocessing these images, including single-rod segmentation and extraction of effective evaluation areas, to optimize image quality. Subsequently, an improved YOLOX model was adopted as the core detection algorithm, with adjustments made to the network structure and loss function to address the characteristics of small target defects. The introduction of a coordinate attention module enables the model to more accurately locate and identify tiny defects. Additionally, the CIoU loss function was employed in place of the traditional IoU loss function to improve the model’s localization precision for small targets. During the model training phase, data augmentation techniques such as Mosaic, Copy and Paste, and Mixup were implemented to enhance the model’s adaptability to new scenarios. The experimental results demonstrate that the improved model excels in the task of end plug defect detection, with significant enhancements in detection accuracy and speed compared to traditional methods and unimproved deep learning models. Tests on industrial datasets show a notable increase in the model’s mean average precision (mAP) while maintaining a fast detection speed, meeting the requirements of actual production. The model also performs well in detecting various types of defects, including accurate identification of porosity, swelling, incomplete welding, tungsten inclusion, and plug abnormalities. In summary, this study successfully develops an efficient and accurate automatic detection method for defects in fuel rod end plugs based on deep learning X-ray imaging. This method not only improves the level of detection automation but also provides strong technical support for the safe management and maintenance of fuel rods. Future research will continue to explore the potential for model optimization to better adapt to a wider range of industrial applications.

AbstractThe meltdown of a fuel rod is a severe accident resulting from the overheating of the reactor core. In the present study, a numerical investigation of this process, focusing on the loss of coolant has been conducted. The objective of this study is to conduct a numerical simulation of transient heat conduction and melting at various points within a typical pressurized water reactor fuel rod. In this analysis, heat conduction in the radial direction of a fuel rod, including the UO2 fuel pellet, the gap, and the zircaloy cladding, is investigated. The FRAPCON steady-state code is employed to calculate the operational parameters of the fuel rod. The calculated parameters, such as coolant and fuel temperatures, fission gas fraction, gap heat transfer coefficient, and burnup, are utilized to evaluate and compare the melting phenomena at different time intervals.In the investigation of the phase change in various parts of the pellet and fuel rod, the explicit finite difference (FD) method is utilized with enthalpy instead of temperature-dependent equations. Finally, the temperature history, phase change, and melting map at different points along the radial and axial directions of the fuel rod during coolant loss and heat transfer coefficient reduction are evaluated based on various operating parameters of the core. To enhance the quality of the results, an uncertainty analysis of effective parameters is conducted.According to this analysis, the heat transfer coefficient of the coolant under accident conditions (0.2 ± 5% kWm−2K−1) and the thermal conductivity of the fuel have the most significant impact on the temperature history and melting process. Highlights include the following:1. The meltdown of a nuclear fuel rod is analyzed under a loss-of-coolant accident.2. The enthalpy formula is discretized by the explicit FD numerical method.3. Effective parameters in melting, such as coolant temperature, burnup, and gap heat transfer coefficient, are obtained by FRAPCON.4. The temperature history, phase changes, and melting map of various radial points within the fuel pellet and cladding along the axial direction of the fuel rod are determined. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Eurasian Physical Technical Journal is the property of E.A. Buketov Karaganda University and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Space power nuclear systems require reactors to have characteristics such as compactness, miniaturization and efficiency. The combination of helium xenon mixture gas and the Brayton cycle is a good choice. Existing research on helium-xenon gas cooled small reactors mostly focuses on the system level, with insufficient consideration for the refinement of individual components. The core is the most critical part of the entire Brayton cycle system, and the thermal safety characteristics of the core are the most important component of the safety characteristics of the reactor system. Therefore, mastering the thermal safety characteristics of the core is particularly important. There are not only two forms of heat transfer inside the core: convective heat transfer and heat conduction, but also radiative heat dissipation. In this paper, the three-dimensional simulation method was adopted to establish a 1∶1 model of the helium-xenon gas cooled small reactor core and analyzed the radiation heat dissipation characteristics inside the core. The Monte Carlo method was used to verify the proposed method of calculating radiation angle coefficients using three-dimensional simulation. The relative error between the two calculation results is less than 1%, proving the accuracy of the three-dimensional simulation method. On this basis, a study was conducted on the radiation heat dissipation characteristics of fuel rods in the core of helium-xenon gas cooled small reactors. The sensitivity analysis on geometric parameters such as fuel rod aspect ratio, fuel rod surface temperature and fuel rod length was conducted, and paid attention to their impact on the radiation heat dissipation characteristics of fuel rods. The results show that as the aspect ratio between fuel rods increases, the radiation angle coefficient also increases. When the length of the fuel rod is less than 100 mm, the length of the fuel rod has a significant impact on the radiation angle coefficient. When the length of the fuel rod is greater than 100 mm, the radiation angle coefficient is not affected by the length of the fuel rod. Finally, a universal empirical formula for the radiation angle coefficient of fuel rods in the core of a helium-xenon gas cooled small reactor was established based on the relationship between the radiation angle coefficient and the radial aspect ratio of fuel rods, the arrangement of fuel rods, and the length of fuel rods. The calculation relative error is less than 8.5%. This paper aims to study the radiation and heat dissipation characteristics of fuel rods and core radiation, understand the laws of internal radiation and heat dissipation, and provides technical support for subsequent research on core thermal safety characteristics.

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

There is a recognized need to discuss the preservation of the coolable geometry of fuel rods during long-term cooling after a loss-of-coolant accident (LOCA). Mechanical load induced by an earthquake may be a primary cause of loss of the coolable geometry after a LOCA. In this research, aiming to confirm the preservation of coolable geometry of fuel rods during an earthquake after a LOCA, we investigated the stress applied to the cladding tube with a rupture opening under a vibration condition via finite element analyses for a fuel assembly that considered the mechanical interaction between a fuel rod and spacer grids. The analyses demonstrated that the stress on the cladding tube increased with the axial holding force of the spacer grid on the fuel rod. The stress concentrating at the periphery of the rupture opening was compared with the strength of the post-LOCA cladding tubes determined via bending tests. This research concludes that the fuel rods are likely to be prevented from fracture due to bending arising from an earthquake during post-LOCA cooling unless the oxidation of the cladding tubes exceeds 15% equivalent clad reacted. [ABSTRACT FROM AUTHOR]

The available literature experimental data on corrosion resistance of traditional and advanced fuel rod cladding materials for water-cooled reactors are summarized. A review of zirconium alloys, which have proven themselves in operation for more than half a century, is presented. As noted, the research work is constantly being carried out to improve zirconium alloys by optimizing their composition, in particular, the amount of tin, niobium, iron and oxygen, as well as development of the new alloys. First of all, the direction of these works is stimulated by stringent nuclear energy requirements, including maximum safety, efficiency and environmental friendliness. At the same time, in the last decade, one of the main goals of researchers around the world is the development of nuclear fuel systems, which tolerate severe accidents. Another trigger for this was the accident in 2011 in Japan at the Fukushima-1 NPP. As the most optimal possible solution, it is considered the surface modification of zirconium alloys by the development of chromium coatings. Such coatings provide an increased corrosion resistance and wear resistance, as well as hydrogen pickup reduced at operating temperatures of the primary coolant and in emergencies. A more radical way to increase the fuel rod cladding accident resistance is to replace the zirconium alloy with another one. The best candidates are FeCrAl alloys and duplex stainless steels (DSS), whose corrosion resistance can be 50 times greater than that of zirconium alloys in loss-of-coolant accident (LOCA) conditions. Unfortunately, under the nominal water-cooled reactor operating conditions, a long-term operation such as claddings will lead to the corrosion product formation and its removal to the coolant followed by their activation and formation of deposits in the core and steam generator. This will certainly entail an increase in the radiation-dose rate from the primary circuit equipment. Considering the traditional and advanced water-cooled reactors claddings, which tolerate severe accident scenarios, an optimized zirconium alloy with chromium coating can be considered as the most advanced one. The corrosion resistance of such claddings is at least five times higher compared to traditional zirconium alloys both under normal operating conditions and severe accidents, and will not cause significant neutron absorption, coolant activation or deposit formation in the primary circuit. [ABSTRACT FROM AUTHOR]

The intricate internal structure of fuel rods results in a non-uniform mass distribution, making it imperative to employ analytical methods for accurate assessment. The study utilizes Euler beam theory to derive the transverse vibration equation for beams with varying mass distribution. The approach involves transforming the non-uniform mass beam into a multi-segment beam with concentrated mass points. Modal function relationships between adjacent uniform segments are established based on continuous conditions at connection points. This transformation leads to the conversion of the variable coefficient differential equation into a nonlinear matrix equation. The Newton- Raphson method is then applied to calculate the characteristic equation and mode shapes, essential for determining natural frequencies. To validate precision, the results obtained are compared with those derived from the finite element method. Furthermore, the developed method is employed to assess the impact of gas plenum location and length on the natural frequency of fuel rods. The proposed methodology serves as a rapid design tool, particularly beneficial during the design phase of fuel rods with non-uniform mass distribution, aiding in configuring structural aspects effectively. [ABSTRACT FROM AUTHOR]

The results of thermal calculations of the part of the fuel assembly of the active zone of the VVER-1000 reactor in the stationary mode of operation make it possible to evaluate the mechanical state of the fuel rod cladding, to understand the influence of reactor control methods on the strength and the design acceptance criteria. The main principles of the evaluation of mechanical characteristics of VVER-1000 fuel rods using the START-3 code are presented. The results of the prediction of the mechanical characteristics of the VVER-1000 fuel rods during the 4-year cycle in stationary mode under normal operating conditions and under their violation are illustrated. The maximum values of stress in the fuel rod in the stationary mode of operation are in the range of 60–80 MPa, which cannot cause depressurization of the fuel rod. [ABSTRACT FROM AUTHOR]

The available literature experimental data on corrosion resistance of traditional and advanced fuel rod cladding materials for water-cooled reactors are summarized. A review of zirconium alloys, which have proven themselves in operation for more than half a century, is presented. As noted, the research work is constantly being carried out to improve zirconium alloys by optimizing their composition, in particular, the amount of tin, niobium, iron and oxygen, as well as development of the new alloys. First of all, the direction of these works is stimulated by stringent nuclear energy requirements, including maximum safety, efficiency and environmental friendliness. At the same time, in the last decade, one of the main goals of researchers around the world is the development of nuclear fuel systems, which tolerate severe accidents. Another trigger for this was the accident in 2011 in Japan at the Fukushima-1 NPP. As the most optimal possible solution, it is considered the surface modification of zirconium alloys by the development of chromium coatings. Such coatings provide an increased corrosion resistance and wear resistance, as well as hydrogen pickup reduced at operating temperatures of the primary coolant and in emergencies. A more radical way to increase the fuel rod cladding accident resistance is to replace the zirconium alloy with another one. The best candidates are FeCrAl alloys and duplex stainless steels (DSS), whose corrosion resistance can be 50 times greater than that of zirconium alloys in loss-of-coolant accident (LOCA) conditions. Unfortunately, under the nominal water-cooled reactor operating conditions, a long-term operation such as claddings will lead to the corrosion product formation and its removal to the coolant followed by their activation and formation of deposits in the core and steam generator. This will certainly entail an increase in the radiation-dose rate from the primary circuit equipment. Considering the traditional and advanced water-cooled reactors claddings, which tolerate severe accident scenarios, an optimized zirconium alloy with chromium coating can be considered as the most advanced one. The corrosion resistance of such claddings is at least five times higher compared to traditional zirconium alloys both under normal operating conditions and severe accidents, and will not cause significant neutron absorption, coolant activation or deposit formation in the primary circuit.

The zirconium alloy cladding of PWR fuel rods which undergo high neutron irradiation during service, will cause significant changes in its microstructure, thereby affecting its macroscopic performance. Therefore, the study of neutron irradiation behavior of zirconium alloy cladding is a focus of nuclear field. However, due to the strong radioactivity of materials after neutron irradiation, relevant experiments must be conducted in a hot cell. Therefore, research on the microstructure of irradiated fuel cladding is a difficult task. In this study, the microstructure of M5TM zirconium alloy cladding material after neutron irradiation was studied by means of transmission electron microscope in the hot cell facility of China Institute of Atomic Energy. The samples were from commercial pressurized water reactor AFA3G type spent fuel rods with burnup of 14 GW·d/tU and 41 GW·d/tU, respectively. A cladding sample with a length of about 10 mm from the fuel rod was cut, and the defueling and chemical cleaning in the hot cell were carried out to obtain a clean cladding sample. Then, mechanical sampling methods was used to prepare a thin slice sample of the cladding with 3 mm diameter. Finally, the electrolytic twin-jet thinning method was used to prepare the cladding transmission electron microscopy observation and analysis sample. In addition, to compare the structural changes of zirconium alloy cladding before and after irradiation, the same method was used to prepare un-irradiated observation and analysis samples of the same material. The observation and analysis results of the un-irradiated and irradiated samples reveal that there are native second phase particles (SPPs) inside the matrix structure of the un-irradiated zirconium alloy cladding, and the overall interior of the matrix is with few nano precipitates and no obvious dislocation structure observed. After irradiation, there is no significant difference in the size and distribution of the native SPPs in the matrix compared to the un-irradiated sample, but significant nano precipitates and high-density dislocation structures appear. As the fuel burnup increases, the size of nano precipitates increases. The similarity of dislocation structures between low and high burnup samples indicates that under the burnup of 14 GW·d/tU, the dislocation structures generated by irradiation in the zirconium alloy cladding basically reach saturation state. The results of selected area electron diffraction (SAED) indicate that although there are some amorphous structures in the native SPPs in the matrix after irradiation, the bcc crystal structure is still the main structure, indicating that the SPPs maintain certain irradiation stability at the burnup of 41 GW·d/tU. In addition, the EDS results of the SPPs indicate that with the increase of fuel burnup, the content of Nb element tends to be depleted. Analysis suggests that after neutron irradiation, the Nb atoms in the SPPs of zirconium alloy expand into the Zr matrix, promoting the precipitation of Nb elements in the form of nano Nb rich phases in the Zr matrix.

The vibration of fuel rods in axial flow is a universally recognized issue within both engineering and academic communities due to its significant importance in ensuring structural safety. This paper aims to thoroughly investigate the stability and nonlinear vibration of a fuel rod subjected to axial flow in a newly designed high temperature gas cooled reactor. Considering the possible presence of thermal expansion and large deformation in practical scenarios, the thermal effect and geometric nonlinearity are modeled using the von Karman equation. By applying Hamilton's principle, we derive the comprehensive governing equation for this fluid-structure interaction system, which incorporates the quadratic nonlinear stiffness. To establish a connection between the fluid and structure aspects, we utilize the Galerkin method to solve the perturbation potential function, while employing mode expansion techniques associated with the structural analysis. Following convergence and validation analyses, we examine the stability of the structure under various conditions in detail, and also investigate the bifurcation behavior concerning the buckling amplitude and flow velocity. The findings from this research enhance the understanding of the underlying physics governing fuel rod behavior in axial flow under severe yet practical conditions, while providing valuable guidance for reactor design.

The operating conditions are analyzed, mechanical characteristics are calculated pertaining to the 4-year cycle fuel rods of VVER-1000. The fuel rod strength characteristics are considered as applied to steady-state and transient conditions of reactor operation.The analysis of the per – fuel rod calculations covering a part of fuel assemblies in the VVER-1000 core under steady-state and several transient operating conditions allows one to assess the mechanical state of fuel claddings, the fulfillment of the design criteria of acceptance and to gain the understanding of the effect produced by the reactor control methods on the strength characteristics.Given are the results of computer modelling the stress-strained condition of present-day fuel rod claddings upon the four-year operation within the VVER-1000 core both under steady-state and transient conditions.The calculations involved transients with the application of different control systems, namely, only the boric system and its combination with mechanical organs of control. The calculations evidence that in terms of the strength criteria these algorithms ensure the tolerable local distortions of power rating fields.

The zirconium alloy cladding of PWR fuel rods which undergo high neutron irradiation during service, will cause significant changes in its microstructure, thereby affecting its macroscopic performance. Therefore, the study of neutron irradiation behavior of zirconium alloy cladding is a focus of nuclear field. However, due to the strong radioactivity of materials after neutron irradiation, relevant experiments must be conducted in a hot cell. Therefore, research on the microstructure of irradiated fuel cladding is a difficult task. In this study, the microstructure of M5TM zirconium alloy cladding material after neutron irradiation was studied by means of transmission electron microscope in the hot cell facility of China Institute of Atomic Energy. The samples were from commercial pressurized water reactor AFA3G type spent fuel rods with burnup of 14 GW·d/tU and 41 GW·d/tU, respectively. A cladding sample with a length of about 10 mm from the fuel rod was cut, and the defueling and chemical cleaning in the hot cell were carried out to obtain a clean cladding sample. Then, mechanical sampling methods was used to prepare a thin slice sample of the cladding with 3 mm diameter. Finally, the electrolytic twin-jet thinning method was used to prepare the cladding transmission electron microscopy observation and analysis sample. In addition, to compare the structural changes of zirconium alloy cladding before and after irradiation, the same method was used to prepare un-irradiated observation and analysis samples of the same material. The observation and analysis results of the un-irradiated and irradiated samples reveal that there are native second phase particles (SPPs) inside the matrix structure of the un-irradiated zirconium alloy cladding, and the overall interior of the matrix is with few nano precipitates and no obvious dislocation structure observed. After irradiation, there is no significant difference in the size and distribution of the native SPPs in the matrix compared to the un-irradiated sample, but significant nano precipitates and high-density dislocation structures appear. As the fuel burnup increases, the size of nano precipitates increases. The similarity of dislocation structures between low and high burnup samples indicates that under the burnup of 14 GW·d/tU, the dislocation structures generated by irradiation in the zirconium alloy cladding basically reach saturation state. The results of selected area electron diffraction (SAED) indicate that although there are some amorphous structures in the native SPPs in the matrix after irradiation, the bcc crystal structure is still the main structure, indicating that the SPPs maintain certain irradiation stability at the burnup of 41 GW·d/tU. In addition, the EDS results of the SPPs indicate that with the increase of fuel burnup, the content of Nb element tends to be depleted. Analysis suggests that after neutron irradiation, the Nb atoms in the SPPs of zirconium alloy expand into the Zr matrix, promoting the precipitation of Nb elements in the form of nano Nb rich phases in the Zr matrix. [ABSTRACT FROM AUTHOR]

Computational and theoretical analysis of the ways to reduce the consumption of natural uranium in VVER-type reactors with an erbium burnable absorber in reusing spent fuel is presented. Calculations are carried out using a simplified model of fuel burnup in a reactor with partial refueling without fuel assembly rearrangement. A comparative analysis of the consumption of natural uranium when using gadolinium and erbium as burnable absorbers for options with uranium and REMIX fuel is carried out. For the options using REMIX fuel, the loss in natural uranium consumption in using erbium compared to the option with gadolinium is shown to diminish by 50%. [ABSTRACT FROM AUTHOR]

BackgroundHeat pipe cooled reactor (HPR) has many characteristics, such as reliability, inherent safety, small volume, modularity, and solid core. The nuclear fuel of solid core is seriously affected by high temperature, strong irradiation, and solid constraint when operating, which affect the heat transfer performance and mechanical properties of the core seriously. The stress and gap heat transfer caused by the contact between monolith and other components change nonlinearly with the increase of burnup, and they influence each other. Therefore, the coupled irradiation-thermal-mechanical behavior of the monolith is a complex multi-physics phenomena.PurposeThis study aims to develop a coupled irradiation-thermal-mechanical model to explore the characteristics of gap variation, heat transfer and mechanics during the lifetime of solid core.MethodsFirst of all, based on the geometric parameter and material of a typical solid core of HPR with fuel rod composed of UO2 pellets and 316 stainless steel cladding, a coupled irradiation-thermal-mechanical model was developed and applied to the finite element multi-physics field analysis software COMSOL. The calculation parameter settings mainly referred to the design parameters of the MegaPower reactor. Then, a thermal conductivity model changing with the increase of burnup for UO2, the gap heat transfer model and mechanical contact were introduced in the gaps in the solid core, and both irradiation-induced deformation effect including densification and fission product swelling, and creep effect of UO2 pellets and 316 stainless steel monolith were taken into account. Finally, the model was applied to calculating the typical HPR and the characteristics of gap variation, heat transfer and mechanics were analyzed.ResultsAnalysis results show that pellet temperature and creep of monolith and cladding increase after complete contact between monolith and cladding. A smaller average number of heat pipes around the fuel rod result in higher temperature and stress distribution in the nearby area, and the cladding in this area has a risk of creep failure during its lifetime caused by internal pressure of the fuel rod and contact pressure between the monolith and cladding.ConclusionsThe gap contact can affect the heat transfer and mechanical properties of the solid core of HPR, and even result in an increase in the risk of cladding failure.

This study aims to improve the measurement accuracy of liquid film thickness using a liquid film sensor with an optical waveguide film (OWF). The measurement principle of employing the OWF is based on the detection of light reflection at the liquid film surface with high spatial resolution. Because the curved surface of the liquid film reflects light and increases measurement error, we propose a signal processing method to remove the error factor in the calculation of the time-averaged thickness. This method requires prediction of the surface curvature, and we numerically investigated the characteristics of the output signal related to the reflected light intensity. The analysis results showed that the effect of the curved surface up to the surface curvature of 5.0 mm−1 was negligible because the liquid film thickness showed good agreement with that of the flat liquid film surface within 7% accuracy. Furthermore, we consider the applicable range of liquid film thicknesses under the operating conditions of boiling water reactors (BWRs). We estimated the surface curvature of the liquid film based on the calculation of the critical Weber number and confirmed that the curvature caused under the BWR operating conditions was covered by the analysis conditions of this study. Therefore, our proposed method for signal processing via the OWF enabled us to improve the measurement accuracy of the time-averaged thickness with respect to the base film thickness by extracting accurate surface curvature data. [ABSTRACT FROM AUTHOR]

The vibration induced by the cooling fluid flow around the fuel rods in the fuel Assembly of nuclear reactors causes the rods to be destroyed and eventually leak due to the fretting wear in the place of contact with their supports for a long time. In this paper, the vibration caused by axial fluid flow around a specific fuel rod under elastic supports is numerically simulated. In this study, the fluid flow is modeled using the Large Eddy Simulation (LES) turbulence model in the FLUENT software. The fluid-structure interaction is also modeled using the ANSYS coupling system. To validate the implemented numerical model, the test results of the reported brass rod vibration similar to the studied problem in this research are used. Due to the long execution time of the two-way fluid-structure interaction simulations with a high grid number, the one-way fluid-structure interaction method is proposed. The results of simulations show that the one-way fluid-structure interaction method can be used in cases where the vibration amplitude of the structure is less than the height of the viscous sub-layer. Also, this method reduces the simulation time by 80%. Finally, the results of the flow-induced vibration simulation of the fuel rod show that the vibration range of the fuel rod will increase by 20 times if the contact of the elastic supports with the rod is lost, which will lead to the intensification of the wear caused by the rod oscillation. Also, the main natural frequency of the rod decreases when the rod loses contact with the supports and falls within the range of the reactor excitation frequency, i.e. 0 to 50 Hz, which should be avoided.

The results of experimental studies into the hydrodynamics of the coolant at the outlet section of the cassette fuel assembly (FA) of the RITM-type reactor of a low-power ground-based nuclear power plant are presented. The purpose of the work is to analyze the distribution of the axial velocity and flow rate of the coolant at the exit from the fuel bundle, in the modernized head of the fuel assembly, near the coolant extraction pipe and the openings of the upper base plate as well as to determine those areas of the fuel bundle from which the coolant flow is most likely to enter the pipe selection to the resistance thermometer. To achieve this goal, experiments were carried out on a research stand with an air working medium on a model of the outlet section of a fuel cassette, which includes an outlet fragment of a fuel bundle with spacer grids, models of an upgraded fuel cassette head, an upper support plate, and a coolant extraction pipe. When studying the flow of the coolant flow in the outlet part of the fuel cassette, the pneumometric method and the method of injection of a contrasting impurity were used. An area covering the entire cross section of the model was chosen as the area under study. The picture of the coolant flow is represented by cartograms of the distribution of its axial velocity and flow rate as well as cartograms of the distribution of the contrasting impurity in the cross section of the experimental model. The results of the experiments can serve as a basis for making engineering decisions when designing new cores of RITM type reactors. The obtained database of experimental data can be used for validation of modern CFD programs and one-dimensional thermal-hydraulic codes used to justify the thermal reliability of cores. [ABSTRACT FROM AUTHOR]

The influence of technological processes and manufacturing of uranium dioxide fuel pellets for fuel elements for experimental fuel assembly (FA-X) which was designed as an alternative fuel for the nuclear research installation (NRI) "Neutron Source Controlled by Electron Accelerator" were investigated. Unlike standard production processes of UO2 pellets, the special feature fabrication process of this nuclear fuel type is production of uranium dioxide powder with enrichment of 4.4 %wt. of 235U achieved by mixing of two batches of powders with different uranium contents: 0.4 %wt. 235U and 19.7%wt. 235U, as well as ensuring the required tolerance of fuel pellets without the use of machining operations. A set of design and process documentation were developed in the R&D Center at NSC KIPT. Experimental stack of fuel pellets, fuel elements and a pilot fuel assembly FA-X were fabricated and designed to be compatible and interchangeable with VVR-M2 fuel assembly adopted as a standard assembly for the first fuel loading at the "Neutron Source Driven by an Electron Accelerator" FA. As opposition to the variant of VVR-M2 fuel assembly which consisted of three fuel rods of tubular shape with dispersion composition UO2‑Al, FA-X accommodates six fuel rods of pin-type with UO2 pellet which located in the zirconium cladding (E110) as the closest analogue of fuel rods of VVER-1000 power reactor. Inside cladding locate a 500 mm high fuel stack which is secured against displacement by a spacer. In the basic variant of FA-X the fuel pellets are made of UO2 with 235U enrichment near 4.4 %wt.

The vibration induced by the cooling fluid flow around the fuel rods in the fuel Assembly of nuclear reactors causes the rods to be destroyed and eventually leak due to the fretting wear in the place of contact with their supports for a long time. In this paper, the vibration caused by axial fluid flow around a specific fuel rod under elastic supports is numerically simulated. In this study, the fluid flow is modeled using the Large Eddy Simulation (LES) turbulence model in the FLUENT software. The fluid-structure interaction is also modeled using the ANSYS coupling system. To validate the implemented numerical model, the test results of the reported brass rod vibration similar to the studied problem in this research are used. Due to the long execution time of the two-way fluid-structure interaction simulations with a high grid number, the one-way fluid-structure interaction method is proposed. The results of simulations show that the one-way fluid-structure interaction method can be used in cases where the vibration amplitude of the structure is less than the height of the viscous sub-layer. Also, this method reduces the simulation time by 80%. Finally, the results of the flow-induced vibration simulation of the fuel rod show that the vibration range of the fuel rod will increase by 20 times if the contact of the elastic supports with the rod is lost, which will lead to the intensification of the wear caused by the rod oscillation. Also, the main natural frequency of the rod decreases when the rod loses contact with the supports and falls within the range of the reactor excitation frequency, i.e. 0 to 50 Hz, which should be avoided. [ABSTRACT FROM AUTHOR]

• A dedicated hydraulic test facility for fuel rod irradiation device was designed and built. • The throttling components that match the design flow rate were selected through hydraulic characteristic testing. • Numerical simulation is an effective method to optimize the throttling elements of irradiation devices. In order to explore the rationality of the design of the fuel rod irradiation test facility, this paper presents a comprehensive study based on theoretical analysis, hydraulic characteristics test, and numerical simulation. First, the design requirements of the fuel rod irradiation test device were analyzed, and the fuel rod hydraulic characteristics test device was designed based on the resistance similarity principle. Next, the orifice plate and throttling components that match the design flow rate were selected through hydraulic characteristic testing to verify the flow adjustment range of the device. Then, the flow field characteristics of the throttling component are studied through numerical simulation, and the distribution of parameters such as speed, pressure, and temperature are obtained. The results of these methods proved the rationality of the design of the fuel rod irradiation test device as well as the throttling component. [ABSTRACT FROM AUTHOR]

A computational and theoretical analysis was made of the use of various burnable absorbers placed in a fuel rod for the maximum reduction of the fraction of the reactivity margin compensated by a liquid system based on a boron absorber for VVER-type reactors with extended campaigns. Various arrangements of fuel rods in fuel assemblies at various contents of burnable absorbers (natural gadolinium and erbium) were considered. The analysis was performed using simplified models of fuel burnup at partial refueling, which represented the core as a periodic lattice of polycells. [ABSTRACT FROM AUTHOR]

In this research the combination of laboratory tests and theoretical formula was used to determine the stiffness coefficients of these spacers that is related to a special fuel assembly. In order to validate the numerical model a spatial modal test set-up has been designed, constructed and set up to perform the modal test. The modal test results indicate that equivalent density of the dummy fuel rod depends on the natural mode of the rod. Then, the spring coefficients of the middle spacers and end constraints have been modified by comparing the results of numerical simulation with the results of modal tests defined in different supporting states of the hollow rod. The simulation results show that in addition to the middle spacers, also the end constraints have a significant effect on the natural frequencies of the rod. The Sensitivity Analysis results show that these coefficients depend on the natural mode of the rod. In addition, the results show that by updating the stiffness coefficients of the middle spacers and end constraints the estimation error of natural frequencies has been significantly reduced and the estimation of natural mode shapes based on the modal assurance criterion has been significantly improved. [ABSTRACT FROM AUTHOR]

Coupling fuel performance and neutronics can help improve the prediction accuracy of fuel rod behavior, which is important for fuel design and performance evaluation. A fuel rod multiphysics coupled system was developed with multiphysics software COMSOL and 3D Monte Carlo neutron transport code RMC. The fuel performance analysis module was built on top of COMSOL with the ability to simulate the fuel behavior in two-dimensional axisymmetric (2D-RZ) or three-dimensional (3D) mode. RMC was innovatively wrapped as a component of COMSOL to communicate with the fuel performance analysis module. The data transferring and the coupling process was maintained using COMSOL’s functionality. Two-way coupling was achieved by mapping power distribution and fast neutron flux fields from RMC to COMSOL and the temperature and coolant density fields from COMSOL to RMC. A fuel rod pin lattice was modeled to demonstrate the coupling. Results show that the calculated power and temperature distributions are reasonable. Considering the flexibility of the coupled system, it can be applied to the performance evaluation of new fuel design.

Zr1 –хUхN nitrides have been synthesized via nitridation of solid solutions of uranium in zirconium. High-temperature saturation of the solid solutions with nitrogen has been shown to yield Zr1 –хUхN ceramics uranium-enriched in their central part. We have presented the concept of a new type of fuel, based on Zr1 –хUхN, for high-temperature gas-cooled reactors. Nuclear calculations of a reactor core demonstrate the feasibility of reaching the critical mass at uranium contents of 10 and 20 wt %. The core size of a conceptual reactor has been determined. We have performed thermophysical calculations of the reactor core and demonstrated the feasibility of obtaining high-grade heat. [ABSTRACT FROM AUTHOR]

The effect of a humid nitrogen atmosphere with an oxygen content of 0.08, 0.48, 1.30 vol % and a temperature of 300°С on the change in the structure of the RBMK-1000 fuel composition vs. the size of the defect in the fuel cladding was investigated. It was demonstrated that in the case of fuel cladding damage with an area of more than 0.8 mm2, SNF storage in a nitrogen atmosphere with an oxygen content of more than 0.5 vol % at a temperature of 300°C brings about a rather rapid oxidation of the fuel composition to U3O8. Thus, dry storage of SNF should be carried out without access to oxygen. [ABSTRACT FROM AUTHOR]

In designing the fuel rod, having a reliable prediction of fuel performance is of high importance in order to comply with safety principles. Various computer codes have been provided for this purpose. Each of these codes uses different mechanical models, numerical and analytical methods. Accordingly, the purpose of the present work is to develop Fortran computer software for mechanical and thermal analysis of fuel rods using numerical methods, especially applying the principle of virtual work in mechanical analysis of fuel rods in steady-state conditions. The finite element method is used to solve the equations. The mechanical analysis includes phenomena such as swelling and fuel density and clad creep. Through these phenomena and simultaneous performance of mechanical and thermal analysis, fuel-clad interaction, stress and strain rate in fuel and clad, fuel center temperature, oxide layer thickness, fuel, and clad temperature distribution during operation is obtained by the code. The results of the code are compared with the results of the analytical method which are available in other research works. Finally, for the VVER1000 reactor, mechanical and thermal analysis of the fuel rod was performed over a 1600-day interval. According to the simulation, the fuel-clad interaction occurs after 1250 days.

Due to the special and difficult conditions for the fuel rods inside the reactor core, inspection and quality control of the fuel after the production at the factory and before loading in the core of the reactor is essential. This process is carried out using various non-destructive tests. Generally, neutron radiography is one of the most important non-destructive tools for fresh fuel inspection. In the present work, neutron computerized tomography (CT) of a depleted TRIGA fuel rod is investigated using a digital detection system which includes a scintillator screen, a mirror, and a CCD camera. The experiments were performed in the INUS neutron radiography facility. The results show that the structure and components of the fuel rod such, as pellets and springs, are well visible by using the neutron tomography technique and it can be utilized for precise investigation of the fuel rod’s structure after production at the factory.

Fuel failures may occur during operation of nuclear power plants. One of the possible and most severe consequences of a fuel failure is that fuel may be washed out from the leaking fuel rod into the coolant. Reliable detection of fuel washout is important for handling of leaking fuel assemblies after irradiation is over. Detection of fuel washout is achievable in the framework of coolant activity evaluation during reactor operation. For this purpose, 134I activity is historically used in WWER power units. However, observed 134I activity may increase during operation even if leaking fuel in the core is absent, and fuel deposits are the only source of the fission products release. The paper describes a criterion which enables to reveal the cases when the increase in 134I activity results from the fuel washout from the leaking fuel rods during operation of the WWER-type reactor. Some examples of applications at nuclear power plants are discussed.

The design and scheme for manufacturing fuel rods metallurgically bonded with ribbed aluminum claddings using hot isostatic pressing and contact-reactive brazing are presented. It is shown that the developed scheme can be used both for production of dispersive fuels and high-density fuels based on uranium alloys. The results of investigations of brazed joints of aluminum cladding with a matrix composition based on aluminum and with samples of E110 alloy through copper and silumin coatings are presented. The results of research of brazed joints of an aluminum cladding with an aluminum-based matrix composition and samples of zirconium alloy E110 made through copper and silumin coating are presented. The strength of brazed joints, composition of diffusion layers formed as a result of contact-reactive brazing in a high vacuum have been determined. The modes of hot isostatic pressing that provide crimping of the ribbed cladding of fuel pellets and rods and obtaining a metallurgical bonding between their surfaces have been defined. It is shown that satisfactory bond strength is provided starting from the temperature of 610 °С. The maximum strength values obtained on the compounds Al-(Al+12% Si)-Zr and Al-Cu-Zr are 57.0 MPa and 55.3 MPa respectively. The fracture of the of aluminum samples joints, obtained with the Cu layer at a temperature of 620 °C, occurs on threaded joints at the strength value of 82 MPa. The results of research of the composition of diffusion layers formed by brazing compounds Al-(Al + 12% Si)-Zr and Al-Cu-Zr are presented. It was established that hot pressing provides the best results for manufacturing of fuel rod dummies in the studied range of modes at a temperature of 630 °C, a pressure of 380 MPa and exposure of 20 minutes.

The swelling, corrosion and high-temperature embrittlement behavior of the fast-neutron sodium-cooled reactor standard and test fuel rod claddings was studied following the operation up to a damaging dose of 55 to 69 dpa. The tested characteristics were found to differ sensitively in conditions similar to irradiation for the claddings of the experimental tube conversion technology. Unlike the standard fuel rod claddings, the test rod claddings were additionally heated in the process of fabrication to homogenize the solid solution at different temperatures and austenitization times. On the whole, this led to an increased cladding resistance contrary the damaging factor of the reactor environment. The positive effect is explained by the influence of carbon and the morphology of swelling-reducing alloying elements, as well as by the nature of the carbide and intermetallide phase precipitation. However, the dispersion of the post-irradiation properties which remained significant and was also earlier observed in the standard rods is explained by potential differences in the heat treatment technology and the irradiation temperature in conditions of a hard-to-control coolant flow velocity. The swelling rate and the in-fuel corrosion depth for the test technology tubes were respectively 0.04 to 0.058%/dpa and 20 to 47 μm; similar values for the test material are 0.036 to 0.056%/dpa and 15 to 35 μm respectively. The short-term mechanical properties of the test fuel rods at a temperature of 600 °C showed a smaller tendency towards high-temperature embrittlement. The dispersion of the properties was caused by the chemical and structural heterogeneity as the result of the tube fabrication.

BackgroundLarge-grained UO2 pellet and coated zirconium cladding are the most practical fuel pellet and cladding material scheme for accident tolerant fuel (ATF) solutions, which are expected to be commercially available in the next few years. However, the existing fuel rod design criteria for commercial pressurized water reactor (PWRs) are based on the material characteristics and application experience of traditional UO2 pellets and zirconium alloy cladding. If the properties of the constituent material of fuel rod change, it is necessary to re-examine the usability of the design criteria based on the corresponding research results.PurposeThe study aims to clarify the potential problems of large-grained UO2 pellet and coated zirconium cladding through the fuel rod design standards, and to provide a reference for the follow-up materials testing.MethodsBased on these two material schemes for ATF on the design criteria, latest research on the material properties of large-grained UO2 pellet and coated zirconium cladding were investigated, and the influence of new ATF fuel rod based on these two schemes on fuel rod design criteria was analyzed.ResultsThe results show that most of the design standards can still be used, and the new ATF fuel rods have certain performance improvements in oxidation and hydrogenation, hence the corresponding design margin is increased. However, the newly added cladding coating may introduce new fuel rod failure mechanisms, including cold spraying and some other coating processes that may damage the zirconium substrate and cause a significant decrease in the fatigue performance, local oxidation after coating damage, etc. Due to the influence of the cladding coating, the equivalent cladding reacted coefficient (ECR) originally used to evaluate the embrittlement level of zirconium alloy cladding is no longer applicable.ConclusionTherefore, in view of these new phenomena, it is necessary to carry out further research work and improve the corresponding design criteria accordingly.

This paper presents the results from numerical investigation of the molten (metal or oxide) fuel relocation during a thermal failure of a fuel rod under conditions close to a severe accident with a power increase in a fast rector. Since experimental investigations needed to determine the regularities of an accident with core destruction cannot be carried out under actual reactor conditions due to safety reasons, getting information on the specifics of molten fuel's movement using mathematical simulation methods is urgent. The methods and approaches used in the considered problem to simulate molten fuel flow and heat transfer with the fuel-rod surface are briefly described. Although the main fuel in reactor units is an oxide fuel at present, the calculations were also performed for metallic (uranium) fuel to evaluate its effect on the development of an accident. In the calculations, account was taken for the different regularities of heat transfer of metal and oxide fuels that resulted from a considerable difference in the Prandtl numbers for melts of these materials. The processes were investigated for the first phase of severe accidents covering the period from the onset of fuel-rod melting to the melt escape from the core center. Fuel-rod melting and fuel flow were simulated for a single fuel rod. [ABSTRACT FROM AUTHOR]

• Anomalous thermal effects were detected during thermal cycling. • Thermal anomalies will lead to the shape change of the shell of Zr–2.5Nb alloy. • Zr–2.5Nb alloy was studied under mechanical cyclic loading by X-ray diffraction. • The c/a elongation of Zr–2.5Nb alloy unit cell during deformation was evaluated. • The preferred change in c/a will cause local stresses and failure. Studies were carried out in this work to establish the influence of thermal cyclic and mechanical cyclic loads that simulated the factors of operational impact on structural stability and crystal lattice dilatation of zirconium alloys. It is shown that during cold cyclic deformation at the first moment of loading the Zr–2.5Nb alloy is characterized by significant instantaneous deformation. The effect of anomalous thermal deformations of the zirconium alloy crystal lattices, which are typical for phase transitions of type I and type II, was detected in the operating temperature range (250–350 °C). The absence of phase transformation in this temperature range allows us to suggest the existence of a causal relationship between the anomalous thermal expansion effects of zirconium crystal lattices with orientation changes and phase transformations in grain boundaries. The performed calculations showed the presence of an anomalous change in the crystallite size distribution at a given temperature. The coincidence that temperatures were correlated with size effects and phenomena such as blistering and flecking has been established. The hypothesis of grain-boundary transformations is indirectly confirmed by the results of the second thermal cycle, during which a second jump of the crystal lattice thermal deformations, not related to structural phase transitions, was detected in the range of 550–650 °C. [ABSTRACT FROM AUTHOR]

Abstract X-ray radiography is widely used as a non-destructive method for defect detection and quality checks of nuclear fuel rod's components in the fabrication process. In this paper, an approach based on Monte Carlo method is proposed to obtain optimum filter and exposure parameters such tube voltage, current and exposure for digital radiography of a nuclear fuel rod including UO 2 pellets with a maximum diameter of 11.9 mm. Monte Carlo N-Particle (MCNP) code was implemented to model an digital X-ray radiography system composed of an industrial X-ray machine and a computed radiography phosphor imaging plate. Three main image quality parameters of contrast, resolution and signal to noise ratio (SNR) were evaluated in various exposure conditions. The results show that using a Cu intermediate filter with a thickness of 0.5 mm and a tube voltage in the range of 400–500kV with a minimum exposure of 20 mA min, can provide the highest quality for imaging the mentioned fuel rod. Moreover, the highest quality can be also achieved for tube voltages in the range of 500–600kV in condition that a minimum exposure of 40 mA min is provided. In addition to obtaining optimum conditions, the mentioned method can propose best exposure conditions proportional to the limitation of laboratory apparatuses. As a case study, the obtained results from simulation were implemented for the existence apparatuses in our laboratory and then an experimental work was carried out. The laboratory apparatuses included an X-ray machine with a peak voltage of 300 kV and a standard phosphor imaging plate. In the obtained experimental image, the details and components of the fuel rods including pellets, springs, Zircaloy clad and also the gap between the pellets were clearly observed. The proposed method can easily be implemented for all types of nuclear fuels with various dimensions and material types. Highlights • A Monte Carlo method proposed to optimize exposure parameters for fuel rod imaging. • Contrast, resolution and SNR were evaluated in various exposures using MCNP code. • Selection of tube voltage, exposure and filter to provide highest image quality was studied. • Experimental imaging of 3 fuel rods were done based on simulation results. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]