Integrity Of Nuclear Power Plants Research Articles

A Novel Ultrasonic Leak Detection System in Nuclear Power Plants Using Rigid Guide Tubes with FCOG and SNR.

Leak detection in nuclear reactor coolant systems is crucial for maintaining the safety and operational integrity of nuclear power plants. Traditional leak detection methods, such as acoustic emission sensors and spectroscopy, face challenges in sensitivity, response time, and accurate leak localization, particularly in complex piping systems. In this study, we propose a novel leak detection approach that incorporates a rigid guide tube into the insulation layer surrounding reactor coolant pipes and combines this with an advanced detection criterion based on Frequency Center of Gravity shifts and Signal-to-Noise Ratio analysis. This dual-method strategy significantly improves the sensitivity and accuracy of leak detection by providing a stable transmission path for ultrasonic signals and enabling robust signal analysis. The rigid guide tube-based system, along with the integrated criteria, addresses several limitations of existing technologies, including the detection of minor leaks and the complexity of installation and maintenance. By enhancing the early detection of leaks and enabling precise localization, this approach contributes to increased reactor safety, reduced downtime, and lower operational costs. Experimental evaluations demonstrate the system's effectiveness, focusing on its potential as a valuable addition to the current array of nuclear power plant maintenance technologies. Future research will focus on optimizing key parameters, such as the threshold frequency shift (Δf) and the number of randomly selected frequencies (N), using machine learning techniques to further enhance the system's accuracy and reliability in various reactor environments.

Numerical investigations on the viscoelastic-damage behaviors of RIVE-induced concrete

Due to the remarkable influence of aggregate characteristics on the mechanical behaviors of concrete materials, and in particular the microcracking development, complex aggregates shapes instead of widely used spherical-shape need to be considered in advanced numerical approaches. Apart from that, in numerical simulations, an extra consideration of the influences of concrete’s heterogeneity on its softening behaviors is strongly recommended by the scientific community. In this context, the softening behaviors of concrete materials undergoing heterogeneous expansion in the aggregates due to fast neutron radiation were investigated in the present paper, and the influences of both aggregate shape and the heterogeneity of aggregate expansion were considered. According to the real shapes of aggregates, polygon-shaped aggregates with random orientation were placed into a 3D numerical concrete specimen, and a two-stage finite element (FE) approach allowing a coupled thermal–mechanical analysis was proposed. In the first stage, the temperature field evolves in time and the 3D concrete specimen was estimated. In the second stage, the temperature- and neutron fluence-dependent heterogeneous expansion in the aggregates was calculated and introduced into the 3D mesoscopic model, and the development of microcracks in the concrete, the resulting mechanical responses, the loss in mechanical properties of the radiation-induced concrete was evaluated. Surmounting the limitations due to the spherical-shape of aggregate and the homogeneous assumption of the aggregate expansion in the concrete, the present paper developed a new methodology in which the combined effects of aggregate’s shape and the heterogeneous distribution of aggregate expansion on the mechanical properties of radiation-induced concrete were considered, permitting evaluating the changes in the dimensional and mechanical properties of radiation-induced concrete. The proposed model provides a reference to assess the long-term durability and the structural integrity of nuclear power plants.

Inverse Heat Conduction Problem in Estimating Nuclear Power Plant Piping Performance

Abstract Most of today's operating nuclear plants that were originally designed for 30 or 40-year life are facing the long-term operation issues. Therefore, it is of meaningful importance to assess the time-dependent degradation and the aging of the relevant nuclear systems, structures, and components because of resulting loss of structural capacity. In this framework, the inverse method is implemented starting from temperatures at an accessible boundary, which are measured through a monitoring system. The reconstruction technique uses the elaborated signal provided by the monitoring system to determine temperature at inaccessible surface: this is the so-called inverse heat transfer problem. The inverse space marching method is applied. Analytical and numerical studies are performed taking into account thermal transient conditions in order to determine thermal loads. In particular, the developed code demonstrates to be able to reconstruct temperature and stress profiles in any section of the pipe with a good accuracy. In addition, the thermal loads obtained suggest that the investigated transient condition is not able to jeopardize the integrity of nuclear power plant, confirming the possibility of the plant extension of life.

Stress-Sensing Method via Laser-Generated Ultrasound Wave Using Candle Soot Nanoparticle Composite.

This article aims to develop a semi-noncontact stress-sensing system using a laser-generated ultrasound (LGU) wave assisted by candle soot nanoparticle (CSNP) composite. While the acoustoelastic effect is commonly targeted to measure the stress level, efforts to combine it with the LGU wave signal have been lacking due to weak signal intensity. In this study, the CSNP-based transducer is designed to potentiate the photoacoustic energy conversion. To demonstrate the wave propagation with the designed parameters, a numerical simulation was first conducted. The experimental results showed that a laser intensity of 6.5 mJ/cm2 was enough to generate the subsurface longitudinal (SSL) wave from the CSNP composite transducer. The normal beam projection is the most effective wave-generation method, exhibiting the highest signal magnitude compared with inclined projection cases. Finally, the laser-assisted stress-sensing system was assessed by increasing the internal pressure of an air tank. The sensitivity of the developed sensor system was estimated to be 0.296 ns/MPa, showing a correlation of 0.983 with the theoretical prediction. The proposed sensing system can be used to monitor the structural integrity of nuclear power plants.

Statistical analysis of parameter estimation of a probabilistic crack initiation model for Alloy 182 weld considering right-censored data and the covariate effect

Abstract To ensure the structural integrity of nuclear power plants, it is essential to predict the lifetime of Alloy 182 weld, which is used for welding in nuclear reactors. The lifetime of Alloy 182 weld is directly related to the crack initiation time. Owing to the large time scatter in most crack initiation tests, a probabilistic model, such as the Weibull distribution, has mainly been adopted for prediction. However, since statistically more advanced methods than current typical methods may be applied, we suggest a statistical procedure for parameter estimation of the crack initiation time of Alloy 182 weld, considering right-censored data and the covariate effect. Furthermore, we suggest a procedure for uncertainty evaluation of the estimators based on the bootstrap method. The suggested statistical procedure can be applied not only to Alloy 182 weld but also to any material degradation data set including right-censored data with covariate effect.

Effects of the Training Transfer Management on the Workers in Nuclear Power Plants

Objective: The aim of this study is to enhance the efficiency of education and training through application and management of "Transfer of Training" in nuclear power plants.<BR><BR> Background: Despite the sophistication and standardization of job-related skills and techniques of workers, accidents/incidents keep taking place due to human errors and unsafe actions and behaviors, which translates into the necessity to review and examine the effectiveness and influence of education and training on the workers of nuclear power plants.<BR><BR> Method/Results: This study drew the factors of "Transfer of Training" through a review on the preceding studies and document research. In addition, through expert examination, this study explored the expected effects and possibility of application when managing the influencing factors of "Transfer of Training" in nuclear power plants. And lastly, management priority order for nuclear power plants was drawn through an AHP analysis.<BR><BR> Conclusion: Among the "Transfer of Training" factors, the training design factor was the most important. In addition, the design of the training and transfer and goal setting showed a high degree of importance among the influencing factors.<BR><BR> Application: The management of "Transfer of Training" in nuclear power plants enhances the capability of workers and improves the operational integrity of nuclear power plants.

重合メッシュ法を用いた溶接熱影響部におけるき裂進展解析

Failure in heat affected zone (HAZ) in welding has been a serious problem for the integrity of nuclear power plant. Trediction of crack behavior affected by the residual stress field in HAZ is important. In this paper, S-Version FEM (S-FEM) is applied to simulate the crack growth under thermal and residual stress fields. S-FEM is applied because the mothod enables us to generate cracked models flexibly. For evaluation of stress intensity factor, virtual crack closure integral method (VCCM) is employed. In order to confirm the validity of this analysis, numerical results are compared with previously-reported analytical and experimental results. Then, crack growth analysis in piping structure with welding joint was conducted. The residual stress data was provided by JAEA, Japan Atomic Energy Agency, based on their numerical simulation. Using S-FEM, two- and three-dimensional analyses are conducted, and crack growth behavior under thermal stress field is studied and discussed.

A Study on Experimental Evaluation for Fracture Behavior of Carbon Steel Pipes for High Pressure Service with Local Wall Thinning

The locally wall thinned phenomenon of pipes is simulated as metal loss due to erosion/corrosion. Therefore, fracture behaviors of pipes with local wall thinning are very important for the integrity of nuclear power plant. In this study, monotonic bending tests without internal pressure are conducted on 1.91-inch diameter Schedule 80 STS370 full-scale carbon steel pipe specimens. We investigated fracture strengths and failure modes of locally wall thinned pipes that welded and unwelded by four point bending test. From test results, we could be divided three types of failure modes.

Evaluation of Wall-Thinning in Pipes Using Laser-Generated Guided Waves

Local wall thinning is one of the major causes for the structural fracture of pipes of nuclear power plants. Therefore, assessment of local wall thinning due to corrosion is an important issue in nondestructive evaluation for the integrity of nuclear power plants. In this study, lasergenerated guided waves were used for pipe inspection, where a laser beam illuminated through linear slit array was used as the transmitter and the air-coupled transducer was used as the receiver. Slits was used in order to enhance the mode-selectivity of guided waves, since the space of slits is equal to the wavelength of the generated wave. The air-coupled transducer detected the selected single mode by turning its detection angle that was calculated from the relations between the wave propagation velocity in air and the phase velocity in dispersion curves. Experimental results for a 4- mm thick carbon steel pipe showed that the detection of the specific mode was useful in the distinction of the wall-thinning thickness in the carbon steel pipe.

Application of Laser-Generated Ultrasound for Evaluation of Thickness Reduction in Carbon Steel Pipes

In carbon steel pipes of nuclear power plants, local wall thinning may result from erosion-corrosion or flow-accelerated corrosion(FAC) damage. Local wall thinning is one of the major causes for the structural fracture of these pipes. Therefore, assessment of local wall thinning due to corrosion is an important issue in nondestructive evaluation for the integrity of nuclear power plants. In this study, laser-generated ultrasound technique was employed to evaluate local wall thinning due to corrosion. Guided waves were generated in the thermoelastic regime using a Q-switched pulsed Nd:YAG laser with a linear slit array. . In this paper, time-frequency analysis of ultrasonic waveforms using wavelet transform allowed the identification of generated guided wave modes by comparison with the theoretical dispersion curves. Modes conversion and group velocity were employed to detect thickness reduction.

감육배관의 굽힘하중에 의한 손상모드와 파괴거동 평가

Fracture behaviors of pipes with local wall thinning are very important for the integrity of nuclear power plant. In pipes of energy plants, sometimes, the local wall thinning may result from severe erosion-corrosion (E/C) damage. However, the effects of local wall thinning on strength and fracture behaviors of piping system were not well studied. In this paper, the monotonic bending tests were performed of full-scale carbon steel pipes with local wall thinning. A monotonic bending load was applied to straight pipe specimens by four-point loading at ambient temperature without internal pressure. From the tests, fracture behaviors and fracture strength of locally thinned pipe were manifested systematically. The observed failure modes were divided into four types; ovalization, crack initiation/growth after ovalization, local buckling and crack initiation/growth after local buckling. Also, the strength and the allowable limit of piping system with local wall thinning were evaluated.

Fracture behavior of straight pipe and elbow with local wall thinning

Fracture behavior of pipes with local wall thinning is very important for the integrity of nuclear power plant. Then we studied the fracture behavior of straight pipe and elbow with local wall thinning. For the straight pipe, failure mode, limit load and allowable wall thinning limit based on plastic deformation ability have been studied systematically. Twenty two straight pipe specimens were tested. The failure mode was divided into four types; cracking, local buckling, ovalization and plastic collapse (ovalization+buckling). Maximum load was successfully evaluated using plastic section modulus and modified flow stress, in dependent to failure mode. For the elbow, plastic collapse and low cycle fatigue fracture by reversed loading have been tested using ten specimens. Observed failure modes were ovalization and local buckling under monotonic loading, and were local buckling and cracking under cyclic loading, especially local buckling promoted crack initiation. Test results were compared with ASME design curve and allowable limit of local wall thinning will be discussed.

Monitor of structural integrity for nuclear power plants

The database management, the load monitoring and the graphical user interface (GUI) are presented. The Combination of database management, load monitoring and GUI forms a useful tool for monitoring the structural integrity of nuclear power plants.

Calculation of stress intensity factors of thermally loaded cracks using the finite element method

The knowledge of stress intensity factors of cracks under conditions of thermal shock loading is very important for the assessment of the integrity of nuclear power plants. Under certain thermal loading conditions, the transient temperature distribution in a component is influenced by the existence of cracks, and thermal singularities may occur at the crack tip. These facts are usually not considered in thermal shock analysis. In this paper stress intensity factors for such load cases are determined by use of the Finite Element (FE) Method. By comparison with analytical results for steady state problems, it is shown that methods for the evaluation of stress intensity factors which are available in nearly every FE code can be successfully applied to problems with thermal singularities. Cracks in plates and in a hollow cylinder subjected to transient thermal loading are discussed. The results are compared with those available from existing FE studies.

原子力発電施設の構造健全性 : 最近の課題と対策

原子力発電施設の構造健全性 : 最近の課題と対策