In-service Inspection Research Articles

Instant testing and non-contact diagnosis for photovoltaic cells using K-means clustering and associated hyperspectral imaging

Renewable energy, particularly solar energy, has experienced remarkable growth in recent years. However, the integrity of solar photovoltaic (PV) cells can degrade over time, necessitating non-destructive testing and evaluation (NDT-NDE) for quality control during production and in-service inspection. Hyperspectral (HS) imaging has emerged as a promising technique for defect identification in PV cells based on their spectral signatures. This study utilizes a HS imager to establish a diffuse reflectance spectra signature for two groups of PV cells: working and non-working. A non-contact photoluminescence imaging-based methodology is employed, using a halogen lamp as an illumination source to replicate sunlight. Our findings reveal that non-working PV regions can be differentiated from working regions within the 400–600 nm wavelength range, with an optimal candidate peak frequency of 450 nm. To accurately group active PV regions in the constructed HS images at 450 nm, we employ an image processing strategy that combines K-means clustering (K-mc) with contour delineation. Specifically, K-mc with K = 8 is used to efficiently and precisely group active PV regions. We demonstrate the effectiveness of this proposed approach and compare it with traditional infrared (IR) imaging techniques. This imaging clustering approach can be implemented using a conventional camera and a 450 nm wavelength filter for NDT-NDE on exterior-mounted PV panels. Overall, the proposed HS imaging technique, coupled with K-mc, offers a rapid and effective means of identifying defects in PV cells, outperforming conventional IR imaging techniques. This advancement contributes to increased efficiency and extended lifespan of solar PV panels.

In-Service Inspection & Repair strategy for newcleo Lead-cooled Fast Reactor

Newcleo is designing and building the next generation of nuclear lead-cooled fast reactors (LFRs) fueled with MOX. A fast-spectrum Small Modular Reactor of 30 MWe will be commissioned in 2030 (the Demonstrator) and a First-Of-Kind 200 MWe a few years later. To this end, newcleo develops an incremental strategy, starting first with the consolidation of mature and promising technologies with its own R&D laboratories and collaborations with major institutes and companies. This step aims to demonstrate the applicability of chosen technologies to a full mock-up of the reactor (the Precursor) available in 2026 for experiments. Inclusion of ISI&R (In Service Inspection and Repair) at the earliest stages of designs is a necessity to achieve a sufficient level of maturity and confidence in the technologies, and to increase safety, availability, and preservation of investments of experimental devices and reactors during their whole life.

Feasibility of Safe Operation of WWER-440-Type Nuclear Power Plants for Up to 60–70 Years

Most WWER-440 reactors are operated over the planned operational lifetime of 30 years. Now, the owners/operators are preparing for the second phase of extending the operation. The paper presents an overview of the most important aspects of ageing of the main components of the WWER-440-type reactors based on the operational and ageing management experiences and primary research efforts supporting the operation. The paper aims to demonstrate that the expectations for the operability of these reactors for up to 60–70 years are realistic. The scope of the review is limited to the ageing of the reactor pressure vessel, vessel internals, and steam generator, which are the lifetime-limiting components besides the containment structure. Some of the corrective actions, improvements of in-service inspections, maintenance, and ageing management are also considered in the paper. It is shown that the selection of materials and unavoidable material degradation processes do not limit the operation of the WWER-440 reactors for up to 60–70 years under the established operational environment and ageing management practice.

Fabricated of the SCC Reference Specimen for Increasing of In-Service Inspection Accuracy in Nuclear Power Plant

To achieve the safety assurance of the nuclear power plants, in-service inspection have were enforced periodic for nondestructive test in weldment. Among them, the ultrasonic evaluation is widely used for the measurement of the defects. Generally, non-destructive inspectors have calibrated a UT machine using the reference specimen in the artificial cracks such as EDM, mechanical notch. As the reflected ultrasonic signal of real crack was detected amplitude value 50% lower than artificial cracks. Therefore, domestic ultrasonic inspectors have a little experience in measuring actual cracks generated in the nuclear power plant field, such as SCC(Stress corrosion cracks, below SCC) and thermal fatigue cracking. Accordingly, in order to improve the inspection accuracy of non-destructive inspectors during operation inspection of nuclear power plant structures, it is necessary to have a lot of experience in inspecting actual cracks.In this study, non-penetration SCC for using in educational specimen of the NDT inspector were fabricated by using the simulated nuclear environment system in base metal of STS 304 pipe. The size of the used pipe were 8 inch(80 sch., thickness 12.7mm). The end of the pipe was used a flange to simulating a fixed environment of the nuclear power plant structure. To accelerate the fabrication of SCC, a corrosive environment was composed by used a blending corrosion solution of the NaOH 1M and Na2S9-H2O 0.1M inside the pipe. The potential of hydrogen of the caustic solution was 14. The fabricated SCC was confirmed for crack size and shape by used UT and RT.

Design of the gripper interlock that engages with the DEMO breeding blanket during remote maintenance

The DEMO breeding blanket (BB) must be replaced during the machine lifetime due to the material degradation caused by the neutron irradiation. The large BB segments can therefore be removed through the upper ports of the vacuum vessel by a remotely operated transporter. The size of these ports is however restricted by the magnetic coils causing some of the BB segments to be accessible only on their extremities. The lifting point of these BB segments therefore is away from their centers of gravity also requiring the transfer of bending moments. A concept of the BB transporter was developed recently [1]. It has the required payload capacity and is capable of carrying out also the tilting maneuvers required to extract the BB segments from the VV. The gripper interlock is the interface to the BB segments and is described in this article including the function of its locking mechanism. It has the tightest space constraints of all BB transporter components, and its design is particularly challenging given the large loads to be transferred. The basic concept of the gripper interlock resembles a massive pin with a diameter of approximately 500 mm that is inserted into a countersunk hole in the backside of the BB segment and then locked by an actuated mechanism. The concept allows on the one hand the transfer of large bending moments. The engagement is on the other hand more challenging as compared to the hook of a conventional crane that is required to transfer vertical loads only. In addition the gripper interlock must be designed according to the rules defined for lifting equipment in nuclear power plants and considering increased requirements regarding qualification and in-service inspection since its failure can cause a load drop with the potential to damage the primary confinement.

Ultrasonic Transducers for In-Service Inspection and Continuous Monitoring in High-Temperature Environments.

The inspection of structures operating at high temperatures is a major challenge in a variety of industries, including the energy and petrochemical industries. Operators are typically performing nondestructive evaluations using ultrasound to monitor component thicknesses during scheduled shutdowns, thereby ensuring safe operation of their plants. However, despite being costly, this calendar-based approach may lead to undetected corrosion, which can potentially result in catastrophic failures. There is therefore a need for ultrasonic transducers designed to withstand permanent exposure to high temperatures, so as to continuously monitor the remnant thicknesses of structures in real time. This paper discusses the design of a heat-resistant ultrasonic transducer based on a piezoelectric element. The piezoelectric material, the electrodes, the backing layer, the wires and the casing are presented in detail from the acoustic and thermal expansion point of view. Four transducers optimized for 3 MHz were manufactured and tested to destruction in different conditions: (1) 72-h temperature steps from room temperature to 750 ∘C, (2) thermal cycles from room temperature to 500 ∘C and (3) 60 days of continuous operation at >550 ∘C. The paper discusses the results, as well as the effect of temperature over time on the properties of the transducer.

Non-Destructive Inspection of High Temperature Piping Combining Ultrasound and Eddy Current Testing

This paper presents an automated Non-Destructive Testing (NDT) system for the in-service inspection of orbital welds on tubular components operating at temperatures as high as 200 °C. The combination of two different NDT methods and respective inspection systems is here proposed to cover the detection of all potential defective weld conditions. The proposed NDT system combines ultrasounds and Eddy current techniques with dedicated approaches for dealing with high temperature conditions. Phased array ultrasound was employed, searching for volumetric defects within the weld bead volume while Eddy currents were used to look for surface and sub-surface cracks. The results from the phased array ultrasound results showed the effectiveness of the cooling mechanisms and that temperature effects on sound attenuation can be easily compensated for up to 200 °C. The Eddy current results showed almost no influence when temperatures were raised up to 300 °C.

Non-Contact MEMS-Sensor Array Inspection of Composites and Metallic Parts Using Lamb Waves

Abstract Composite materials are becoming more popular in the aerospace industry, because of their physical properties. In quality assurance and in-service inspection, there is a need for fast, non-contact, high-quality, non-destructive inspection techniques. The most common approach is to perform the inspection using water-coupled high-frequency transducers. Full wavefield techniques are promising to replace the conventional inspection approach. However, these are currently performed by a laser vibrometer setup, which has drawbacks. As an alternative, a low-cost micro-electro-mechanical system (MEMS) sensor array and dedicated processing scheme are presented enabling fast inspection of large samples. This inspection approach uses a piezoelectric actuator to excite the composite or metallic part with Lamb waves. An array of MEMS sensors records the energy that radiates into the surrounding air. A dedicated processing scheme will translate the measured wavefield into a thickness map of the inspected part. For composite parts, the material’s anisotropy needs to be taken into account for accurate thickness mapping. In principle, all relevant defects show up as local thickness reductions. The results in this paper are obtained with a MEMS-sensor array of 128 elements capable of detecting ultrasound up to 250 kHz at a typical stand-off distance of 100 mm. Defects up to 6 mm in diameter could be detected in thick panels, and defects as small as 2.5 mm could be detected in thin panels. A full-size fuselage experiment shows that the method is also suited for fast inspection of large inspection areas.

Tolerance of oxide wedging crack growth of manufacturing flaws in CANDU Zr-2.5Nb pressure tubes to cyclic operating stresses

In-service inspections of Zr-2.5Nb pressure tubes in CANDU nuclear reactors have detected inner diameter surface breaking flaw indications which are classified as manufacturing flaws. It is considered prudent to evaluate potential growth by means of a postulated oxide wedging crack growth mechanism. An oxide wedging crack growth model has previously been developed and applied to predict future flaw sizes. Fatigue crack growth due to cyclic operating stresses is calculated for these flaws, and this is added to the postulated oxide wedging crack growth for fitness-for-service evaluations. This established approach is overly conservative since fatigue crack growth cannot occur while the oxide wedging crack growth mechanism is active, due to the essentially displacement-controlled loading of the flaw by the oxide wedge, which causes the stress intensity factor for the flaw to be effectively constant under cyclic operating stresses. This paper addresses this conservatism by evaluating crack growth due to postulated oxide wedging, and demonstrating that fatigue crack growth is not possible under oxide wedge loading.

Higher Order Mode Cluster ( HOMC) Guided Waves - A novel technique for accurate detection and quantification of hidden corrosion

Corrosion happening at the support regions, under sleeves, in interfaces of the pipes are severe and need reliable assessment. These hidden corrosion vary in nature based on the type of hidden region such as in resting supports, welded sacrificial pads or clamp supports just pipes wrapped in protective wrapping. Ultrasonic NDT plays a major role in the inspection of these supports. Several NDE Techniques exist for such inspection but quantifying the wall loss in an accurate manner depends on the nature and extent of corrosion. A new concept for the improved inspection of corrosion at the support region of pipelines using a short-range ultrasonic guided wave technique that uses a collection of Higher Order Modes Clusters, called here as HOMC* has been developed and verified at several field sites. Ultrasonic guided waves (circumferential or axial modes), once generated will be reflected or transmitted from corrosion and other features on the pipe. Inspection can be carried out from the accessible portion of the pipe . This could be done as an in-service inspection without emptying the pipe. HOMC guided waves are unaffected by the point of incidence of the probe. Accurate Wall loss estimation is possible together with cross sectional profile of maximum thickness loss. Using this information an accurate remaining wall thickness assessment is achieved. The technique has been deployed from pipes 2inches and above and upto 20mm wall thickness.

Wall thickness measurement and defect detection in ductile iron pipe structures using laser ultrasonic and improved variational mode decomposition

A reliable non-destructive evaluation technique that accurately assesses the wall thickness and slag inclusions of ductile iron pipes prior to annealing is crucial to reducing costs in subsequent operations. The rough outer layer of the produced ductile iron pipes leads to considerable background noise, limiting the probability of detection of laser ultrasonic testing equipment. In this study, an improved variational mode decomposition approach based on particle swarm optimization is suggested initially to increase the signal-to-noise ratio. This technique develops a fitness function suitable for processing ultrasound signals. A scanning laser ultrasound technique is designed for simultaneous measuring wall thickness and finding slag inclusions in ductile iron pipes based on excitation locations and flight time. A finite element analysis simulation model is established to assess the feasibility of this scanning method. This experiment is conducted on a 4.2 mm thick sample with slag inclusion defects and a 6.2 mm thick sample without slag inclusion flaws. Experimental results demonstrate that the developed denoising technique may enhance the average signal-to-noise ratio from 34.83 dB to 43.15 dB; the accuracies of thicknesses measurement for two ductile iron pipes are 96.90% and 99.52%, respectively, and the slag inclusion defects can be recognized by ultrasonic amplitude or energy distribution in different B-scan images. It has potential implications for in-service inspection of ductile iron pipe structures.

Microstructure-based polycrystalline finite element modeling of Inconel X-750 irradiated in a CANDU reactor

Aging of Inconel X-750 spacers in Canada Deuterium Uranium (CANDU) reactors is managed through a combined program of in-service inspection, component removal for material surveillance testing, research and development, and fitness-for-service evaluation. Crush tests of ex-service spacer coils have shown that the load carrying capacities of Inconel X-750 spacers decrease with operating time. The spacers also exhibit reduced ductility and inter-granular failure that is typical of helium (He)-embrittlement. Examination of ex-service spacers confirmed that material interfaces (mainly grain boundaries) were perforated by He-bubbles. Assuming that the grain boundary bubble coverage dictates the grain boundary strength and thus the failure load of the polycrystalline component, we have simulated the degradation and inter-granular failure of the spacer coils at low and high operating temperatures.The observed inter-granular fracture has been simulated by a cohesive zone model that governs the grain boundary de-bonding process. The grain boundary strength is related to the perforation by He-stabilised cavities. The model of grain boundary traction separation consists of three parts: (i) a rate-theory calculation to predict grain boundary coverage; (ii) a local material degradation model to quantify helium induced grain boundary strength reduction; (iii) a traction separation model to simulate grain boundary de-cohesion. The results show that the accumulation of He-stabilised cavities on grain boundaries causes material failure at lower stresses and strains that are dependent on the irradiation temperature and the neutron exposure. The load carrying capacity of the X-750 spacers decreases non-linearly with irradiation and the rate of change is decreasing with increasing dose.

Research on the on-line fracture assessment of cracked primary loop pipes in NPPs

The cracked pipes usually need to be replaced or conservatively evaluated in the Nuclear Power Plants (NPPs). There is still a lack of effective methods for a real-time crack propagation and safety evaluation based on massive monitoring transient data. In order to ensure the safe and economic operation of the NPPs, the authors have developed an on-line fracture assessment technology in this paper. The calculation method of total stress suitable for on-line monitoring data, optimizing the calculation method of transient thermal stress and stress intensity factor (SIF) based on the influence function method are discussed. Then, the on-line software for the primary loop hot-leg pipe which is based on the research results is developed. In this paper, an on-line safety assessment method based on the monitoring data which can meet the requirements of the in-service inspection code (RSE-M Code) is given. The influence function is proposed to be established at a fixed time interval which will make the computational complexity reduced significantly. By the parallel calculation method and the characteristics of the algorithm, the origin algorithm along the time axis is optimized to an algorithm that can be calculated to obtain the target parameters at each time point within the specified attenuation time range of the influence function. Then the calculation efficiency is significantly improved. By compiling the data and calculation methods in the code, the calculation method of SIF suitable for on-line monitoring is obtained, and the reliability is also verified by the analyzed case. Based on the technical results obtained from the research and the requirements in the RSE-M Code, the re-development is carried out in the general commercial software platform. The software is developed, and it is demonstrated that it has sufficient analysis accuracy.

Study on In-Service Inspection of Nuclear Fuel Assembly Failure Using Ultrasonic Plate Wave.

As protection for nuclear power plants is quite necessary, the nuclear fuel is sealed in zirconium alloy thin wall cladding. During service, fuel rods might be damaged caused by wall-thickness thinning, cladding corrosion and cracking, etc. This will cause the coolant to enter into the fuel rod, which may lead to the failure of the fuel assembly. However, current diagnostic methods have limitations due to the special structure of the fuel assembly and the underwater and radioactive environment. In this paper, a novel inspection method is proposed to recognize the failure of a fuel rod. The fuel rod failure can be detected based on the presence or absence of coolant inside the fuel rod by using an ultrasonic plate wave. The inspection model and process algorithm are proposed for in-service inspection. The relationship between signal and scanning position is established and analyzed. Both ultrasound field simulation and experiment have been carried out for validation. The corresponding results illustrate that the failed nuclear fuel rod of the whole fuel assembly (including the internal rods) can be effectively detected without the influence of the near-field region by using the proposed method.

Efficient Numerical Integration Algorithm of Probabilistic Risk Assessment for Aero-Engine Rotors Considering In-Service Inspection Uncertainties

Numerical integration methods have the characteristics of high efficiency and precision, making them attractive for aero-engine probabilistic risk assessment and design optimization of an inspection plan. One factor that makes the numerical integration method a suitable approach to in-service inspection uncertainties is the explicit derivation of the integration formula and integration domains. This explicit derivation ensures accurate characterization of a multivariable system’s failure risk evolution mechanism. This study develops an efficient numerical integration algorithm for probabilistic risk assessment considering in-service inspection uncertainties. The principle of probability conservation is applied to the transformation of the integration domain from the current flight cycle to the initial (N = 0) computational space. Consequently, the integration formula of failure probability is deduced, and a detailed mathematical demonstration of the proposed method is provided. An actual compressor disk is evaluated using the efficient numerical integration algorithm and the Monte Carlo simulation to validate the accuracy and efficiency of the proposed method. Results show that the time cost of the proposed algorithm is dozens of times lower than that of the Monte Carlo simulation, with a maximum relative error of 5%. Thus, the efficient numerical integration algorithm can be applied to failure analysis in the airworthiness design of commercial aero-engine components.

Probabilistic risk assessment for the piping of a nuclear power plant: Uncertainty and sensitivity analysis by using SINTAP procedure

For the purpose of probabilistic fracture mechanics estimation in the field of reliability analysis of pressure vessels and piping, a probabilistic method by using SINTAP (Structural INTegrity Assessment Procedures for European Industry) is developed. Monte Carlo simulation (MCS) is applied to investigate the influence of both the uncertainties of random parameters and inspection strategies on the failure probability (FP). A sensitivity analysis with regard to the above-mentioned aspects is conducted to identify the most significant parameters and efficient countermeasures to reduce failure risk in specially-assumed cases. The results of this study point out that the thickness and the inner radius of the pipe have the strongest impact on the FP of the pipe, followed by the bending moment, weld residual stress (WRS), crack length, initial crack depth, yield strength (YS), ultimate tensile strength (UTS), flow strength, fracture toughness, inner pressure and axial force. Compared with the detection quality and the inspection start time, selecting an appropriate inspection interval is a more effective method to reduce the failure risk. All these results provide a reference for the design of piping systems and guide the selection of in-service inspection (ISI) strategies with suitable flaw-detection capabilities and inspection frequencies.

Allowable stresses related to crack growth stability and leak-before-break for circumferentially cracked pipes governed by limit loads: Part I, Pipes subjected to tensile loading

When cracks are found in high toughness ductile pipes governed by limit loads at in-service inspections, and the applied stresses at the cracked locations are less than the allowable stresses, the cracks are allowable. The allowable stresses are generally determined in combination with failure stresses and safety factors. Failure of pipes with circumferential part-through cracks subjected to tensile loading occurs in a two-step sequence: first there is the onset of crack penetration, and then there is the break after the crack penetration. The Local Approach of Limit Load Criteria, which is based on the net-section stress approach, can estimate both crack penetration and break stresses for high-toughness ductile pipes. Using the Local Approach of Limit Load Criteria, allowable tensile stresses based on crack penetration stresses are determined by means of safety factors for plant service level conditions. The allowable stresses decrease with increasing crack angles and depths. The characters of the allowable tensile stresses are found to vary with changing sizes of crack angles and depths. These different characters involve the presence/absence of two parameters: leak-before-break (LBB) and crack growth stability. The crack size boundary on LBB vs non-LBB can be expressed by appropriate equations containing crack depths and angles. In addition, the boundaries of crack growth stability/instability are also expressed by crack depths and angles. These boundaries are taken into consideration during non-destructive inspection and crack growth evaluation.

Prediction of the in-reactor deformation of Zr-2.5wt%Nb pressure tubes using the crystal plasticity finite element method framework

To predict the irradiation induced deformation of the pressure tube material in Canada Deuterium Uranium (CANDU) reactors, a deformation model based on the crystal plasticity finite element method framework was developed in this study. Based on the principle of superimposition, three deformation mechanisms—thermal creep, irradiation growth, and irradiation creep—were incorporated into the model. The model parameters were optimized and verified using in-service inspection (ISI) data from Korean CANDU reactors by considering the irradiation effect on deformation mechanisms and the influence of incompatible grain interactions. We found that the model with optimized parameters could successfully describe the results of not only the pressure tube ISI data but also the simulated mini capsules experiments in a test reactor. By analyzing the optimized irradiation creep parameters and results, the possibility of diffusion-base mechanisms was elucidated and the anisotropy in the activation energy terms along the tube directions was confirmed. For predictions using the model, three strategies for extracting the optimized parameters from available measurement data were examined. As a result, we found that, the better prediction ability can be achieved when only the latest data measured at the same reactor before the predictions are used to optimize the model parameters (method 3).

Radioactive Waste Embedding in Cement Matrix and the Establishment of Cemented Waste Testing Laboratory at Paks Nuclear Power Plant

For the treatment of low and intermediate level radioactive wastes generated at the Nuclear Power Plant of Paks, a new compact waste package will be created with a new cementation technology currently under construction. During the operation, the solid waste will be placed in steel drums, which then will be put into a thin-walled steel container. The cavity volume of the containers will be filled with cemented slurry made using liquid radioactive waste. The compact waste packages must comply with the waste acceptance criteria described in the National Radioactive Waste Repository Safety Report. The task of the cemented waste testing laboratory, which is under construction, will be the in-service inspection of the finished cemented material and the determination of properties such as compressive strength and leachability (diffusion coefficient) of the specimens made thereof. During the installation of the laboratory and testing of the devices used in the cement / concrete industry, it turned out that in many cases special moulds are required due to the low viscosity and corrosive properties of the radioactive paste. In addition, the collection of debris generated during the measurement of compressive strength of radioactive cement stone and the minimization of dusting are important requirements.

International Context Regarding Application Of Single Failure Criterion (SCF) For New Reactors

The paper provides an overview of the regulatory design requirements for new reactors addressing Single Failure Criterion (SFC) in accordance to international best-practices, particularly considering the SCF relation to in-service testing, maintenance, repair, inspection and monitoring of systems, structures and components important to safety. The report [1] discusses the detailed comparison of the current SFC requirements and guidelines published by the IAEA, WENRA, EUR and nuclear regulators in the United States, United Kingdom, Russia, Korea, Japan, China and Finland. However, this paper presents the summary of work from [1] and 2major examples from IAEA and WENRA and applications for small and modular reactors.