Zirconium Tubes Research Articles

Fabrication of a new core–shell structured magnetic abrasive particles with good performance

The traditional Fe-based magnetic abrasive powders (MAPs) are prone to degrade by corrosion and oxidation. The extremely low stability to some extent limits the development of magnetic finishing. In this study, a new kind of 410 stainless steel-based spherical MAPs with core–shell structure was synthesized by gas nitriding. The nitride layer shows the dendrite structure containing (Fe, Cr)4N phase and lath-like CrN phase. The Zirconium tube roughness can be reduced from 0.280 to 0.153 μm. The origin of the good finishing performance was investigated through the Vickers hardness and nanoindentation tests.

On the contribution of local anisotropic creep to macroscopic irradiation-induced growth in zirconium alloy polycrystals

ABSTRACT Zirconium alloys used in nuclear reactor exhibit, under fast neutron flux, a macroscopic deformation even without applied stress called irradiation induced growth. Because of the polycrystalline nature of the material, the local growth of individual grains results in strain incompatibilities yielding to intergranular stresses and thus to local creep. In order to study and understand the influence of the local anisotropic creep on the macroscopic growth behaviour, an analytical and a numerical study have been undertaken, using Voigt and self-consistent estimates and also fast Fourier transform simulations. It is shown that the anisotropic local creep has a strong influence on the effective macroscopic growth strain of the polycrystal. Especially, when the deformation is difficult along the 〈 c 〉 axis, a growth enhancement effect is observed. This phenomenon is well explained in the frame of the Voigt estimate using a reduced fibre texture. Computations conducted using a texture representative of the industrial material provide a quantitative confirmation of this enhancement effect. This work demonstrates the significant contribution of the local anisotropic creep to the macroscopic in-reactor growth strain of zirconium alloys. Highlights The local creep anisotropy has a strong influence on macroscopic irradiation growth of polycrystalline zirconium tubes. The macroscopic axial growth can be significantly enhanced by the interactions between grains when local creep deformation is prevented along the 〈 c 〉 axis. A simplified representation of grain interactions (reduced fibre texture and Voigt estimate) offers a qualitative understanding of this phenomenon. Refined estimates of the effective growth strain are provided for realistic crystallographic texture using polycrystalline unit-cell FFT computations.

Achieving in-situ Fe-based spherical particles for magnetic abrasive finishing process of Zirconium tube via gas nitriding method

The magnetic abrasive powders have great influence on the magnetic abrasive finishing accuracy and efficiency. A new kind of Fe-based spherical magnetic abrasive particles with Fe2N precipitates and Fe4N matrix was successfully synthesized by gas nitriding. The results show that the prepared Fe-based magnetic abrasive particles can reduce the roughness of the Zirconium tube from 0.325 to 0.154 μm by four MAF passes. The high hardness and wear resistance of the Fe-based magnetic abrasive particles are assessed based on the Vickers hardness and nanoindentation results. This study can provide theoretical and experimental support for fabricating high-performance magnetic abrasive particles.

Co-optimization of magnetic abrasive finishing behaviors of zirconium tube surfaces with Fe-6.5 wt% Si/SiC abrasives using BP neural network and response surface methodology

A novel abrasive of Fe-6.5 wt%Si/SiC composite was prepared for magnetic abrasive finishing (MAF) to achieve fine machining of zirconium alloy tubes. The optimized parameters of magnetic pole speed R= 16.0 r·min−1, tube speed Q= 1220.5 r·min−1, abrasive quality P = 119.3 g, and machining gap N = 2.30 mm were obtained by response surface analysis. The BP neural network model was used to predict the optimal machining effect of MAF. It showed that the predicted results of BP neural network model are quite close to the actual test results, with an error of 7.526%. Finally, the roughness of zirconium alloy tube was reduced from 0.207 µm to 0.113 µm, resulting in a 45.41% improvement rate. Fe-6.5 wt% Si/SiC composite abrasive was an efficient magnetic abrasive to improve the surface quality of zirconium alloy parts.

Application of Artificial Intelligence for Automated Defect Detection on Nuclear Power Plant Fuel Asssembly Components

For manufacturing nuclear fuel, Framatome's various plants and workshops carry out rolling operations of chemical and thermal treatment on zirconium tubes. After the finishing stages of these tubes, a series of checks are carried out, including a visual inspection of the final external surface, to ensure the quality and integrity of these cladding tubes, essential for nuclear safety. Today, this visual inspection is automated but requires a significant amount of time for an expert to review the rejected components by the current analysis system. To minimize the number of components rejected by the existing system while respecting the same level of defect detections, a study of the application of Artificial Intelligence was conducted. Two convolutional networks of a similar architecture have been developed: one for making a decision to sanction the tube and one to classify the type of rejection. In this article we present the architecture of the chosen neural networks, as well as the performance obtained for a large set of evaluation data from the rejections of produced tubes. Finally, we put these results into perspective by exposing the potential productivity gain using this system in production assisted by Artificial Intelligence.

A precision grinding technology for zirconium alloy tubes based on ultrasonic wall thickness automatic measurement system

To ensure the safety and long-term performance of nuclear fuel cladding zirconium tubes, the wall thickness uniformity of each cross-section is strictly needed. Therefore, this paper developed an automatic ultrasonic wall thickness measurement system for detecting and precision grinding the nuclear zirconium tubes. The overall scheme, key mechanical structures, and design of the control system are determined considering the project background and measuring requirements. Then, a series of performance testing was performed from aspects of measuring accuracy, measuring efficiency, stability, and practicability. The results revealed that it could accurately obtain the wall thickness distribution with high accuracy and efficiency, and subsequently the precision grinding process by automatically generated deviation correcting procedures to achieve the requirement of the wall thickness uniformity and surface quality. The proposed new combination method of ultrasonic auto-measuring and precision grinding proposed in this work would have great significance for the development and application of nuclear reaction zirconium alloy containers.

INFLUENCE OF HYDROGEN SATURATION ON CHARACTERISTICS OF ZIRCONIUM

The article presents a literature review on the effect of hydrogen saturation on the properties of zirconium alloys. Zirconium alloys are an indispensable structural material for the core of nuclear reactors. During operational loads, the interaction of zirconium materials with interstitial elements (oxygen, nitrogen, hydrogen) necessarily occurs. Zirconium shows the greatest affinity for hydrogen. One of the vulnerable zirconium elements is fuel rod tubes. The safe operation of a nuclear reactor depends on their integrity. In addition, it is fuel tubes that are most exposed to hydrogen. Therefore, the systematization of knowledge about the effect of hydrogen on the properties of zirconium tubes will make it possible to better predict their operational behavior. According to scientific literature data, depending on the volume of absorbed hydrogen, either a solid solution or zirconium hydrides can form. The dependence of the absorbed hydrogen zirconium on the dilution of the hydrogen medium and temperature is shown. The chemical composition of the zirconium alloy also affects the rate and amount of absorbed hydrogen. The effect of hydrogen on the mechanical properties of zirconium alloys is presented. The differences on the fracture surface after tensile tests at room temperature are shown depending on the amount of absorbed hydrogen. Data are presented that indicate that hydrogen atoms are located in octahedral or tetrahedral interstitial voids of a hexagonal close-packed zirconium lattice. It is shown that the thermal solubility of hydrogen in α-zirconium is extremely low, its value is ~6 at. % at the eutectoid transformation temperature, and at room temperature the solubility of α-Zr hydrogen does not exceed 1·10-5 wt. %. In high-temperature β-Zr, hydrogen dissolves up to ~50 at. %. It has been established that deterioration of the properties of zirconium elements of nuclear reactors during operation due to exposure to hydrogen is likely due to a number of factors: hydrogen embrittlement, the formation of large massive accumulations of hydrides and delayed hydride cracking. It is shown that the direction of arrangement of hydrides depends on the texture of the matrix and on the stresses present in the material that act during the formation of hydrides. It has been established that hydrogen can penetrate into a metal through an oxide film, diffusing, for example, along extended defects such as dislocations and grain boundaries. It is noted that the solubility of hydrogen in zirconium depends on other penetration elements already present in the metal, for example, the solubility of hydrogen in α-zirconium depends on the soluble oxygen in the metal matrix, which is confirmed by the reduced «Zr–O–H» ternary system.

Achieving in-situ alloy-hardening core-shell structured carbonyl iron powders for magnetic abrasive finishing

The finishing accuracy and efficiency of magnetic abrasive finishing (MAF) are mainly depending on magnetic abrasive powders (MAPs). A new kind of core–shell structured carbonyl iron powders (CI-MAPs) with a hard Fe/Al intermetallic shell for magnetic abrasive finishing is successfully synthesized by in-situ alloy-hardening the surface of spherical carbonyl iron powders in this study. The feasibility of such an in-situ alloy-hardening strategy is theoretically designed according to the Fe/Al intermetallic compound formation chemical reaction using the Gibbs free energy principle and the TG-DSC testing, and thus the experimental thermodynamics temperature and time ranges of such chemical reactions are proposed. Experimental results shown that a densely zigzag-like uniform alloy-hardening layer with a thickness of about 11 μm, which are composed of Fe3Al, FeAl, and Fe2Al5 intermetallic compounds, was in-situ chemically synthesized on the surface of the spherical carbonyl Fe powders. The achieved core–shell structured powder is performed to finishing a Zirconium tube experimentally, and the roughness (Ra) of the Zirconium tube is greatly improved from 0.361 μm to 0.085 μm by 3 MAF passes.

Phase Transformation and Atypical Variants in an Extruded Two Phase Zirconium Tube

The texture evolution of a Zr-2.5wt pctNb alloy pressure tube is characterized after the $$\alpha \rightarrow \beta $$ transformation and the $$\beta \rightarrow \alpha $$ transformation by room temperature EBSD maps and reconstruction of the high temperature $$\beta $$ phase. The $$\beta $$ phase forms a deformation texture which varies along the radial direction of the tube and changes after thermal cycling. Variant selection during the $$\beta \rightarrow \alpha $$ phase transformation is shown to be strongly influenced by residual stress within the material, with a more randomized variant selection for water quenched samples. Slower cooling rates were found to create orientations in addition to the 12 $$\alpha $$ variants expected by the Burger’s orientation relationship in a given $$\beta $$ grain.

Electrowinning of Nuclear-Grade Zr from Ba2ZrF8–ZrF4 Salt System

The electrowinning of nuclear-grade Zr from a Ba2ZrF8–ZrF4 salt system at 780°C was investigated by using a zirconium tube and a copper plate as cathode materials. An equimolar Ba2ZrF8+ ZrF4 mixture was used as a low-melting-point electrolyte. Here, Ba2ZrF8 is a single source of zirconium and ZrF4 is an auxiliary product to convert Ba2ZrF8 into a low-melting-temperature BaZrF6 salt (Tmelt = 730°C). The cyclic voltammetry measurement indicated that the reduction of Zr4+ ions has a two-step character corresponding to the Zr4+/Zr2+ and Zr2+/Zr0 transitions. The electrochemical reduction behavior of Zr ions on the zirconium tube (inert cathode) and the copper plate (reactive cathode) was analyzed in terms of deposition time and the phase composition of the deposited layers. The metal deposit obtained on the Zr electrode was a single-phase Zr metal, whereas on the Cu electrode, two layers consisting of Cu–Zr and Zr phases were revealed. The further melting of the as-prepared Zr deposit yielded a nuclear-grade Zr ingot with 99.99% purity.

Результаты моделирования процесса штамповки ячеек дистанционирующей решетки из тонкостенных циркониевых труб

Результаты моделирования процесса штамповки ячеек дистанционирующей решетки из тонкостенных циркониевых труб

Ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure at 350 °C

Zirconium alloy is used as nuclear fuel cladding in nuclear reactors because it offers the merits of low neutron absorption cross-section, high corrosion resistance, excellent mechanical properties, and satisfactory creep resistance under operating conditions. However, the cladding experiences uniaxial and multiaxial stresses, which can lead to material failure due to the ratcheting effect. In this context, in this study, a series of uniaxial and multiaxial ratcheting tests were conducted on zirconium alloy tubes at 350 °C. The experimental results show that the temperature significantly affects the properties of zirconium alloy, and its ductility at 350 °C is poorer than that at room temperature. The axial ratcheting strain rate greatly increases with increase in the mean stress, stress amplitude, and peak stress holding time. In addition, the loading history also greatly affects the uniaxial and multiaxial ratcheting of zirconium tubes. A lower stress level after a loading history with higher stress level leads to shakedown of the ratcheting, while a higher mean stress after a loading history with lower mean stress brings down the ratcheting strain rate. The existence of a constant internal pressure significantly retards the failure of the zirconium tubes.

State of the RBMK CPS Channels from Post-Reactor Studies

Serviceability validation of the efficiency of CPS channels operating beyond the design basis service life is one of the tasks in reactor service life extension problems. Periodic post-reactor examinations of sections of a channel in shielded enclosures are conducted as part of routine monitoring. Currently, CPS channels with operating times 4–37 calendar years have been investigated. It was found that according to the shape change, corrosion state, hydrogen content, mechanical properties, and characteristics of the fracture toughness of zirconium tubes, electron-beam weld connections, and steel-zirconium adapters the CPS channels remain serviceable. An increase in hydrogen content was found in the material of the tubes and adapter nipples of the CPS channels after operation for more than 26 years. The high hydrogen content is accompanied by the appearance of hydride accumulations on the inner surfaces of local regions. The nature and causes of this phenomenon have not been found.

Microstructures and Properties of Typical Zirconia Tubes for the Determination of Oxygen Content

The measurement of oxygen content in liquid iron and steel has been a challenge for decades and there have been many attempts to develop oxygen sensor to adjustment the smelting process on the basis of the probe for monitoring oxygen dissolved in molten liquid. However, there was still not a common standards of zirconium tubes about domestic product, which limited their application and improvement This work selected several domestic enterprises oxygen measuring half-cell system to study the relationship between the physical properties and microstructure at room temperature. Furthermore, we also analyzed about the typical performance of the domestic Oxygen content products, which would be beneficial to test and improve the quality of domestic products. Keywords: zirconia tubes, oxygen content, oxygen sensor

Modeling of critical conditions in a fuel cell of a nuclear reactor based on combustion of energetic SHS systems

It is shown that critical conditions (temperature and pressure) in a fuel cell (metal tube piece) of a model nuclear reactor can be produced by burning energetic thermite SHS systems. The heating temperature of the zirconium or steel tube can be controlled by varying the composition of the thermite mixture, the gap between the thermite pellet and the tube wall, and the design of the thermite pellet. The results of the studies show that a WO3-2Al based thermite mixture can be used to model an emergency situation (without the use of radioactive fuel) in the reactor fuel cell.

Effects of different surfaces on the transport and deposition of ruthenium oxides in high temperature air

In order to understand the behaviour of ruthenium oxides in the reactor coolant system during an air ingress accident, new tests were performed in the frame of the RUSET (RUthenium Separate Effect Test) experimental program. These aimed to ascertain the effects of different surfaces (quartz, stainless steel (SS), zirconium alloy, alumina, oxidised metal, and surfaces with Mo or Cs deposits) on the transport and decomposition of ruthenium oxides in air stream along the temperature gradient zone (1100–100 °C). The results demonstrated that the heterogeneous phase decomposition of RuO 3 and RuO 4 to RuO 2 is catalysed more efficiently by the quartz surface than by the SS or alumina surfaces. The presence of MoO 3 layers decreased the RuO x precipitation extent on all investigated surfaces. The trapping effect of Cs deposit on Ru in the temperature gradient zone was proved in the case of the SS surface. On the contrary, presence of Cs precipitate on alumina and especially on quartz surfaces was found to decrease their catalytic effect on the decomposition of ruthenium oxides, and thus increased the RuO 4 concentration in the outlet air. Similarly to the effect observed for Cs deposition, the presence of other fission products in the evaporation area (at 1100 °C) decreased the partial pressure of RuO 4 in the outlet air at the SS surface and increased it at quartz and alumina surfaces. When zirconium (E110) cladding material was placed in the temperature gradient zone, no Ru transmittance occurred until the high temperature end of the zirconium tube was completely oxidised. After the intense oxidation of E110, Ru release occurred only in the presence of other fission product species. Pre-oxidation of SS surfaces in steam had no significant effect on the Ru passage.

Numerical Simulation of Residual Stress and Deformation of Special Resistance Welding of Nuclear Power Zirconium Tube

According to the actual situation of special resistance welding capp of nuclear power zirconium tubes, using nonlinear finite element analysis technology to establish an three-dimensional numerical analysis model. Analysising the distribution rules of welding residual stress and deformation. Simulation results can be explained a number of welding-related stress problems and also provide a favorable reference for optimizing the welding capp method and implementation of capp processes.

Computational-experimental studies of the neutronics of the SM reactor with different variants of the neutron-trap arrangement

The results of experimental studies of the neutronics of the high-flux SM reactor with different arrangements of the neutron trap are presented. The MCU series of high-precision computer programs implementing the Monte Carlo method is used for computations. Experimental data on reactivity effects, the effectiveness of safety and control rods, and the coefficients of nonuniformity of energy release in the core have been obtained in experiments on a critical assembly ‐ a physical model of the SM reactor ‐ and directly in experiments in the reactor. The error is 4.2‐10% in determining the reactivity parameters and 5‐10% for the relative energy release in the fuel elements. Information on the neutron field formed in the volume of the neutron trap has been obtained for two arrangements of the beryllium and water moderators. The differential and integral energy spectra of the neutrons in the energy interval from 0.5 eV to 20 MeV are obtained for three points inside the trap (external, central series, center). The flux density of thermal, superthemal, and fast neutrons are determined. The SM high-flux research reactor was designed to produce radionuclides and test reactor materials, fuel compositions, and fuel elements [1]. The results of the studies performed with this reactor are used to substantiate the serviceability and safety of nuclear reactors; the radionuclides produced find applications in industry and medicine. The SM reactor has been operated with three different arrangements of the neutron trap. In 1961‐1990, an experimental channel which was connected to an autonomous secondary loop with light water as the coolant was located at the center of the reactor. The next phase of the construction of the reactor, the purpose of which was to increase operational safety, was completed in 1991‐1992 [2]. To increase 252 Cf production, the radionuclide in greatest demand at the time, the water neutron trap was replaced with a centralized beryllium block for holding transuranium targets. The coolant in the primary loop of the reactor started to remove heat from the central block. The cross section of the rebuilt SM reactor is shown in Fig. 1. In 2002, the central beryllium block in the central moderating cavity of the reactor was replaced with a separator structure consisting of 27 14-mm in diameter and 0.5-mm thick zirconium tubes with water in the space between the tubes. The purpose of the change in the construction was to increase production efficiency of 63 Ni, 75 Se, and other isotopes which were in demand at the time and to accumulate which a low-energy neutron spectrum with high flux density is required. The cross section of the central moderating cavity of the SM reactor with different arrangements of the neutron trap is shown in Fig. 2.

Experimental Investigation and Elastic-Plastic Analysis on Connection Strength of Zirconium Tube-Tubesheet Joints

Zirconium tube is widely used in heat exchanger equipments in petrochemical process for significant corrosion resistance. The connection joint of tube-tubesheet is the weakest parts in a heat exchanger. The experiment and numerical analysis of different materials (zirconium tubes, titanium tubes and 16MnR tubesheets, 316L tubesheet) joints were performed in this paper. The expansion joints specimens were prepared at the pressure ranging from 28MPa to 38MPa. And pulling out test was performed from 20°C to 300°C. The finite element model of tube-tubesheet joint was established. The effect of expansion pressure, temperature and groove on the pulling out strength of joints was analyzed. Both the experiments and the finite element analysis show that the pull-out strength increases with the increasing expansion pressures. Working temperature also has a great effect on the connection strength of tube-to-tubesheet joints, especially for the zirconium and 316L joints, which have the most significant difference of thermal expansion coefficient between tube and tubesheet. The residual contacting pressure on the contact surface between tubes and the tubesheet is not uniformly distributed and two tightness bands are found near the surfaces of the tubesheet or at the two brinks of the groove on the tubesheet hole. Compared with the ungrooved joint, the residual contacting pressure on the tightness bands for the grooved joint is much higher, indicating a grooved joint has better tightness.

Extracting quantitative data from partly revealed anisotropic microstructures as applied to zirconium tubes

Pressure tubes holding the uranium rods of a CANDU (CANada Deuterium Uranium) reactor show deformation with time in service. At a certain point, this deformation becomes too severe and the tubes must be decommissioned. This reduces the reactor’s energy output. It has been observed that the microstructure of the pressure tubes is indicative of creep behavior. This paper presents the image analysis procedure used to characterize the level of anisotropy and the α-phase dimension of the Zr-2.5Nb alloy used in pressure tubes. The techniques used to evaluate experimental error and to correct for sampling bias are also explained.