Nuclear Cladding Tube Research Articles

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.

Modeling and tests on the damage and fracture behaviors of non-irradiated zirconium alloys with different hydrogen concentrations at RT and 350°C

Zirconium alloys are widely used as the materials of nuclear cladding tubes or some other structural components. The chemical reaction of zirconium with coolant water leads to hydrogen pickup during reactor exposure, degrading the mechanical properties of zirconium alloy tubes. In this study, the load-displacement curves and fracture morphologies of ring-typed un-irradiated zirconium alloy specimens with various hydrogen contents are obtained at RT and 350 °C by circumferential tensile tests. An efficient modeling strategy for the damage and fracture behaviors of hydrided zirconium alloy specimens is proposed and numerically implemented, reproducing the local necking, damage initiation and evolution, crack propagation and ultimate fracture of the specimens. The predictions of load-displacement curves for different hydrided specimens agree well with the experimental data, and the fracture morphologies are also captured. The combinations of experimental tests and numerical simulation herein successfully identify the mechanical constitutive models of hydrided zirconium alloy materials, well describing their hydrogen-concentration related characteristics of strain hardening, softening and collapsing rupture. In the considered mean hydrogen concentrations less than 800 ppm, the hydride-induced hardening and embrittlement are predicted to appear dominantly at room temperature (RT); these effects are negligible at 350 °C, resulting from the greatly improved ductility of hydrides and zirconium alloy matrix at high temperatures. This work lays a foundation for simulation of the failure behaviors of hydrided zirconium alloy tubes.

Microstructural evaluation of gas nitrided AISI 316 LN austenitic stainless steel

Direct utilization of austenitic stainless steel is limited in numerous industries due to its low surface hardness. To improve the hardness of steels, gas nitriding is a convenient technique to enhance the surface hardness of austenitic stainless steel (316 LN). AISI 316 LN austenitic grade stainless steel were utilized for various applications like nuclear cladding tubes, heaters, turbine parts and blower parts. Nitiriding was carried out for three different hours namely 14, 24 and 72 hrs for the austenitic stainless in the current study. A maximum surface hardness of 1200 HV is obtained for gas nitirding technique when compared to 210 HV for the untreated specimen. Gas nitrided specimen shows a uniform nitride layer on the 316 LN stainless steel. Similarly XRD analysis of the nitrided specimen unveil the distribution of iron nitride phase which occurred during nitriding process. An effective case depth of the compounded layer from gas nitriding was found to be 30.7 µm.

Determining failure properties of as-received and hydrided unirradiated Zircaloy-4 from ring compression tests

An integrated experimental and computational approach was developed to determine the plastic strain at fracture and to investigate the fracture energy of as-received and hydrided unirradiated Zircaloy-4 nuclear cladding tubes in the hoop direction at room temperature from ring compression testing (RCT). This work builds on previous methods in the literature that were developed to obtain the mechanical properties of nuclear cladding before fracture (Young’s modulus, yield stress, and strain-hardening parameters) from experimental RCT results and extends the characterization of this material class to include material failure properties. A ductile damage approach was developed and implemented in a validated finite element model to predict material failure, evaluate fracture energy, and quantify the plastic strain at fracture initiation. The hydrogen content of the Zircaloy-4 specimens was varied up to 610 wppm when the failure became brittle. As expected, increasing hydrogen content caused embrittlement of the Zircaloy-4 samples tested and the plastic strain at fracture and fracture energy to decrease

Optical measurement of the high temperature ballooning of nuclear fuel claddings.

During the last decades, several experiments revealed interesting details about the ballooning and burst of nuclear cladding tubes in accident conditions involving high temperature and high pressure. A new set of experiments was conducted at MTA EK to further improve our understanding of this phenomenon. The research goal was to determine the pressure of the burst at different temperatures. These data represent a durability and usability limit of fuel cladding tubes under loss of coolant accident conditions. A large tube furnace was fitted with two optical telescopes on opposite sides to observe the ballooning and burst of zirconium alloy fuel cladding samples. These telescopes gave an opportunity to capture the ballooning and also to use a high-speed camera to record details of the crack formation and crack propagation during the burst. A precise pressurization system was constructed to feed high-pressure argon into the samples at a given pressurization rate. The previous data were reproduced accurately in this new apparatus with further insight into the phenomena. The change in tube diameter was observed using regular HD cameras. Based on the captured images, an algorithmic method was developed to separate the contours of the cladding tubes from the background and the diameter increase was measured. It was found to be mostly uniform at lower pressures. The formation of a bulge started at 85% of actual burst pressure. Crack formation at high temperature was recorded using a high-speed camera at speeds up to 100.000 fps. The details of the planning, facility construction, and gauge calibration are detailed in this article.

Optical observation of the ballooning and burst of E110 and E110G cladding tubes

Abstract During the last decades several experiments revealed interesting details of the ballooning and burst of nuclear cladding tubes in accident conditions involving high temperature and high pressure. To further investigate this phenomenon, a new set of experiments was conducted at MTA EK. A large tube furnace was fitted with two optical telescopes on opposite sides to observe the ballooning and burst of Russian E110 and E110G (sponge based E110) fuel cladding samples. The experiments were recorded using regular and high-speed cameras. Based on the captured images we developed a method to separate the contours of the cladding tubes during the ballooning to measure the change in diameter. Two successive modes of ballooning were observed, the uniform growth of the samples was followed by an asymmetrical, local ballooning (bulge formation) 2 s before the burst. It was found that the samples have bent by approximately 5° relative to their original axis in the latter phase and this was not caused by the jet effect of the high pressure argon gas escaping after the burst. Every sample opened up on the convex side of the bend. Axial grooves have formed on the surface of the E110G samples under tension. The high-speed camera was used to capture the cracking and the burst. Prior to the burst, a high temperature spot was observed at the position where the crack would initiate. The timescale of the crack propagation was 0.2 ms, the crack tip was estimated to be at least 100 °C hotter than the rest of the sample. The burst pressure was determined between 700 °C and 900 °C at different pressurization rates.

A radial heat flow apparatus for thermal conductivity characterisation of cylindrical samples

This paper presents a recent implementation of the radial heat flow method, a classical technique used to measure the thermal conductivity of materials, including the hardware designed to this end, as well as the theoretical formalism used to analyse the data. The novelty is in its application to high-aspect-ratio hollow cylindrical samples, more specifically sections of nuclear cladding tube, whereby samples representative of full length fuel clads can be measured instead of platelets, like in usual laser-flash measurements. The apparatus designed and built to this end can accommodate tubular samples of varying length, diameter and wall thicknesses made of most materials, with the specific goal of determining the through-thickness thermal conductivity of silicon carbide composites, a highly anisotropic material under consideration as a fuel cladding material.

Effect of Heat Treatment on the Creep Properties of Zr-1Nb-0.12O Nuclear Cladding Tubes

Creep properties of one-step and two-step annealed Zr-1Nb-0.12O cladding tubes were studied. Creep tests were carried out at 450~500 °C with the applied stress between 80MPa and 120MPa The creep rates of the two-step annealed Zr-1Nb-0.12O alloy were found to be slower than those of the one-step annealed Zr-1Nb-0.12O alloy. The creep rate decreased with increase of grain size with annealing for Zr-1Nb-0.12O at intermediate temperatures, suggesting the creep resistance can be enhanced by the grain size control. The creep life of two-step annealed Zr-1Nb-0.12O increased over the one-step annealed Zr-1Nb-0.12O by the factor of 18~20 despite the greater initial instantaneous strain.

Mechanical performance of oxidized Zr-Nb-O nuclear cladding tubes

Ring compression and tensile tests on oxidized Zr-Nb-O cladding tubes were performed, to examine the reliability of nuclear cladding tubes after oxidation. The oxidation rate was observed to be far greater at 700°C than 600°C, because of the increased volume fraction of less protective porous oxide. The tensile strength of oxidized Zr-1Nb cladding tubes at 600°C for 3 h increased, with no appreciable loss of ductility. After heat treatment at 600°C for 24 h and 700°C for 3 h, the yield strength and the initial flow stress increased, and the flow stress decreased rapidly with strain, resulting in decreased ductility. The increase of yield strength after heat treatment at 600°C in Zr-1Nb was associated with the presence of strong and protective oxide film. In compressive loading, for cladding tubes oxidized at 600°C for 24 h and 700°C for 3 h, a small drop of load, resulting from cracking of the surface oxide layer, was observed at the total displacement of 1.3 ∼ 1.5 mm. The catastrophic fracture that was observed at the total displacement of 5.7 mm in Zr-Nb-Sn-Fe did not occur in Zr-Nb-O. The absence of sudden drop and catastrophic fracture at the displacement of 5 ∼ 7 is thought to be associated with the softer matrix of annealed Zr-Nb-O.

Mechanical Reliability of Oxidized Zircaloy-4 Alloy Nuclear Cladding Tube in Compression

Ring compression tests on the oxidized Zircaloy-4 cladding tubes were performed to examine the mechanical reliability of Zircaloy-4 cladding tubes. In compression of the non-oxidized cladding tubes, large load drop associated with the cladding failure appeared at the total displacement of 6.3 mm. After oxidation at 600°C for 24 hrs and 700°C for 3 hrs, small drop of load resulting from small cracks in the thin oxide film was observed at the total displacement of 1.0 mm. After oxidation at 700°C for 24hrs, load was extremely low, but catastrophic fracture occurred at the total displacement of 2.4~2.8 mm. The oxide of Zircaloy-4 cladding tubes formed at high temperature apparently causes the premature failure of tubes.

Mechanical Reliability of Oxidized Zr-1.1Nb-0.5Cu Alloy Nuclear Cladding Tube in Compression

Ring compression tests on the oxidized Zr-1.1Nb-0.05Cu cladding tubes were performed to examine the mechanical reliability of Zr-Nb-Cu cladding tubes. In compression of the non-oxidized cladding tubes, large load drop associated with the cladding failure appeared at the total displacement of 6.3 mm. After oxidation at 600°C for 24 hrs and 700°C for 3 hrs, small drop of load resulting from small cracks in the thin oxide film was observed at the total displacement of 1.0 mm. After oxidation at 700°C for 24hrs, load was extremely low, but catastrophic fracture occurred at the total displacement of 6.2~6.5 mm. The oxide of Zr-1.1Nb-0.05Cu cladding tubes formed at high temperature apparently causes the premature failure of tubes.

Mechanical Reliability of Oxidized Zr–1Nb–1Sn–0.1Fe Alloy Nuclear Cladding Tubes

Mechanical Reliability of Oxidized Zr–1Nb–1Sn–0.1Fe Alloy Nuclear Cladding Tubes

Mechanical Properties of Oxidized Zr-1Nb-0.7Sn-0.1Fe Alloy Nuclear Cladding Tubes at High Temperatures

The ductility loss and the failure of cladding tubes due to oxidation are considered as a threat to the safety of the nuclear. In this study, oxidation behaviors and the effect of oxide layer on the mechanical properties of Zr-1Nb-0.7Sn-0.1Fe alloy cladding tubes with a lower tin content (<1 wt. %) were investigated. The drastic increase of the oxidation rate at 700oC suggests the change in the nature of the oxide layer with increase of temperature. The acceleration of the oxidation rate for Zr-1Nb-0.7Sn-0.1Fe alloy tubes at 700oC suggests the increased volume fraction of less protective porous oxide, supported by the presence of oxide cracks even before mechanical testing. The ductility increased after oxidation at 600oC, suggesting softening due to recovery and recrystallization is predominant over the ductility loss caused by oxidation. The ductility began to decrease significantly after oxidation at 700oC and cladding tubes exhibited the brittle fracture.

The Effects of Wear Debris under Fluid Flow Environment on Fretting Wear Mechanism of Nuclear Fuel Cladding Tube Supported by Supporting Grid

Nuclear cladding tubes are subjected to low-amplitude flow-induced vibrations (FIVs). Fretting wear occurs when two contacting surfaces are exposed to very low oscillation. Because nuclear cladding tubes are damaged by this fretting wear, it is necessary to try to evaluate the lifetime and improve the wear resistance of the tube. For this purpose, it is very important to clearly understand the damage mechanism of the tubes. Therefore, the present study undertook fretting-wear tests to investigate how the fluid that flows between the tube and the supporting grid affects the fretting wear. The governing damage mechanism consisted of abrasive wear and surface detachment behavior. In the case of the fluid flow environment, most of the wear debris was ejected from the contacting surfaces and this mode had a large effect on the maximum wear depth of the tube or wear amount. Additionally, the tests were performed under the water flow state as well as the water/air flow state. The result is that if the fluid flow exists, severe wear occurs through some mechanisms, such as two-body abrasion; further, the fluid flow decreases the contact load and attenuates the kinetics of oxidation. Because the severe wear due to two-body abrasion interrupted the formation of oxide film, wear amount was increased. Furthermore, even in the case of water/air flow environment, the worn surface was hardly oxidized due to extremely severe wear.

Nonsmooth fracture dynamics of functionally graded materials

This paper presents a numerical framework for the simulation of dynamic fracture of heterogeneous material. It consists in a multibody approach based both on the concept of Frictional Cohesive Zone Model and on NonSmooth Contact Dynamics. The heterogeneities of the material are taken into account using a multiscale method. The microscopic scale corresponds to the scale of heterogeneities. The macroscopic scale corresponds to the structure where gradients of properties will be invoked. The ability of the framework is illustrated by the fracture of hydrided Zircaloy-4, constituting nuclear cladding tube, under transient loading.

A New Approach to Measuring Transverse Properties of Structural Tubing by a Ring Test

Abstract A new approach to the ring test is presented in this paper. Three-dimensional elastoplastic finite element modeling with contact has been performed to analyze the stress and strain distribution in the ring, to optimize the ring testing system, and to investigate the effect of friction between the ring specimen and the fixture. Based on the numerical results, a new design of a holding device that creates a uniaxially stressed zone is proposed in order to determine transverse behavior of tubular products, such as the modulus and the stress-strain curve. A case study is presented on nuclear cladding tubes in Zircaloy. The approach shows promise for testing various kinds of materials in structural tubing, including ductile or brittle materials, metals, composites, or polymers.

Status of eddy current testing in India

The status of eddy current testing (ECT) in India from various considerations is discussed in this paper. The points considered are: (i) the industrial back ground of India which has motivated the development and use of ECT; (ii) ECT in quality assurance; (iii) ECT in codes and standards as practised in India; (iv) capabilities of Indian-made ECT equipment; (v) manpower training in ECT in India; and (vi) research and development in ECT in India, including applications such as nuclear cladding tubes, surface probes, measurement of level of liquid metals, composites, and inservice inspection. Future trends expected to develop from the present status are also explained.