Nuclear Power Plant Structures Research Articles

Experimental and numerical examination of low cycle fatigue behaviour on AISI304L steel

The integrity of the components of nuclear power plants must be guaranteed under operating and emergency conditions. It is necessary to evaluate all possible degradation mechanisms that could impact the structural integrity of nuclear power plant structures such as the pipelines and other components of the cooling systems. Environmental fatigue significantly influences the degradation mechanisms of steel components operating in water, which eventually affects the operational lifetime of the components. Exploring non-codified methods for more precise assessment of fatigue phenomena can pave the way for developing safer nuclear power plants with longer operational lifetimes. This is very important as the global demand for cleaner energy is increasing. This research study deals with a numerical investigation of the low-cycle fatigue behaviour of steel used for those nuclear reactor components where environmental fatigue plays a major role. The numerical simulation methodology is proposed to study the dependence of the kinematic hardening parameter on the strain amplitude to investigate the low-cycle fatigue behaviour of AISI 304L steel for the prediction of the fatigue curves that can be used for the estimation of nuclear power plant safety and its lifetime.The experimental data was used to estimate the parameters required to define the material model for the numerical simulation and to validate the results of the numerical simulation of the low cycle fatigue behaviour. The presented methodology can be used for fully reversed constant-amplitude strain loading low cycle fatigue simulations for various strain ranges to predict the low-cycle fatigue behaviour of steel under repetitive loading.

Seismic fragility analysis of nuclear containment structure with Bayesian cloud analysis framework

Nuclear containment structure plays a critical role in safe operation of nuclear power plants. Seismic safety of nuclear power plant structures has become a hot issue in nuclear engineering community following Fukushima nuclear accident. Seismic fragility analysis, as a key element in probabilistic safety assessment of nuclear power plants, is widely used in nuclear power plant structures. This study presents an innovative method for seismic fragility analysis of engineering structures. The proposed methodology combines conventional Cloud analysis method and Bayesian framework. The accuracy and efficiency of the proposed methodology were demonstrated by a single-degree-of-freedom system, heavy computational efforts but high accuracy seismic fragility analysis method was adopted for comparison. Seismic fragility analysis of a typical prestressed concrete nuclear containment structure was analyzed using the proposed method. Results indicated that method proposed in this study can improve the accuracy of seismic fragility analysis. Furthermore, with the enhance of performance level, the increasing ratios for median seismic capacity of the nuclear containment structure was decreasing.

Soil-structure interaction analysis of nuclear power plant structure using multi-step method in time-domain

Recently, several countries have experienced strong earthquakes, and there has been growing interest in the seismic design of structures. Particularly, consideration of soil-structure interaction (SSI) has become important when performing seismic response analysis for nuclear power plant (NPP) structures. A seismic analysis method that considers SSI in the time domain has been proposed in previous studies. This method is applicable to the seismic response of structures, because it can consider the influence of the SSI effect with the impedance function in the frequency domain and simultaneously considers the structural nonlinearity. However, it has only been applied to the structure comprising a single beam-stick. In this study, a modified method for applying it to more general structures is proposed, and its applicability is validated with an actual reactor containment building (RCB) structure comprising a three beam-sticks. Validation was performed by comparing the seismic responses with those of ACS SASSI. Using the proposed method, SSI analysis was performed with three soil profiles and the results were compared to investigate the effect of SSI on seismic responses, including the in-structure response spectrum (ISRS) and time history acceleration. For more realistic analysis, soil profile and earthquake input motion occurred in South Korea are adopted. Further, a multi-step method considering nonlinear behavior was also performed and compared with the linear case.

Seismic Fragility Assessment of Equipment in Nuclear Power Plant Structures Considering Nonlinear Hysteretic Behavior of Squat Walls

Accurate seismic risk assessment of structures necessitates precise fragility analysis. This study focuses on the seismic fragility assessment of equipment in nuclear power plant structures, particularly emphasizing the influence of nonlinear hysteretic behavior exhibited by squat walls. By developing and comparing linear and nonlinear models of reactor containment and auxiliary buildings within a nuclear power plant, this research elucidates the significant impact of nonlinear behaviors on the seismic response and subsequent fragility curves of associated equipment and systems. Utilizing a lumped-mass stick model, the study efficiently captures the characteristics of squat shear walls, facilitating the calculation of floor response spectra and fragility for nuclear facility structures. The impacts of pinching and degradation, inherent characteristics of a squat wall’s hysteresis under cyclic loading, were identified as significant. These include the shift and change in amplitude of the floor response’s peak, along with amplification in high-frequency domains. Fragility assessments, informed by seismic response analysis, indicate significant variations in fragility curve parameters based on equipment location and frequency. The findings question the effectiveness of traditional response factor methods and general regression techniques in accurately addressing the complex probabilistic seismic demands of equipment, thus highlighting the importance of using direct analysis for calculating fragility in situations with significant nonlinearity.

Study on seismic performance of SRC frame-bent structures in CAP1400 nuclear power plants

With the conventional island turbine plant project of a CAP1400 nuclear power plant (NPP) as the research background, this paper proposed to apply steel reinforced concrete (SRC) frame-bent structure system to the main building of the NPP. Substructures comprising three frames and two spans from the prototype structure were identified for testing, and a test model was constructed at a scale ratio of 1/7 to test the dynamic and pseudo-dynamic characteristics of the substructures. On the basis of the test results, the structure was further verified using the finite element analysis method. The results showed that the first three modes of substructure were respectively lateral translation with torsion, longitudinal translation with torsion and torsion; The ratio between the third-order and first-order mode cycles was 0.72, indicating that the torsional stiffness of the structure was relatively small and the torsional effect was obvious. When the maximum acceleration peak was 2200 gal, the maximum inter-story drift ratio of the model structure reached 1/31. Ultimately, when the structure was finally destroyed, the concrete at the bottom of each SRC column was crushed and peeled off.

Seismic performance enhancement of small modular reactors using optimal design of inerter-based vibration absorbers

Nuclear power plant (NPP) structures face significant challenges in ensuring structural resilience subjected to earthquake motions. The inerter-based vibration absorbers (IVAs) have been proven to control structural vibration, but they have not been applied in small modular reactor (SMR) structures. In this study, the optimal IVA parameters are investigated for SMR structures, which focus on reducing the seismic response of the first floor and fifth floor. By using a genetic algorithm (GA), the IVAs parameters are optimized to minimize the storey displacement variance and storey acceleration variance. Besides, the robustness analysis is performed to ensure the optimized parameters are practical. The various ground motions are selected for time history analysis (THA) to illustrate the seismic response of SMR equipped with the optimal IVAs. The results of THA demonstrate that the optimal IVAs enhance the seismic performance of the SMR compared with traditional SMR structures. By positioning optimal IVAs, significant mitigation of both storey displacement and storey acceleration are achieved. This study explores to a deeper understanding of IVAs' efficacy in SMR structures, which is beneficial to the practical use of IVAs in SMR structures.

Site effect on seismic response of buried nuclear power plant structure

The development of multifunctional, multipurpose small modular next generation nuclear reactors has become a trend in designing small-scale nuclear power stations. In this study, taking a new next-generation buried nuclear power structure as the target structure, a 3D finite element model of the site-nuclear power plant structure system was established, and its response at different sites was studied. Using the internal substructure method, this study initially explored the impact of earthquake input and artificial boundary locations on the structural response. The results indicate that the internal substructure method enhances computational efficiency without compromising accuracy. For this model, it is recommended to input earthquake motion close to the structure, set the lateral boundary location to three times the width of the structure, and set the bottom boundary location to two times the depth of the structure. Subsequently, an investigation into the response of the structure at varying site wave velocities is conducted. The results show that as the wave velocity increases, the acceleration response intensifies, while the displacement response diminishes. However, the relationship is nonlinear; the response exhibits less variation with increasing site wave velocity. The seismic response of a structure is profoundly influenced by site conditions, emphasizing the imperative need for careful consideration of specific site characteristics in seismic analyses. In addition, the response pattern of the overall structure is significantly different from that of conventional large surface-sited nuclear power plants, and further research should be conducted on the seismic response characteristics of buried nuclear power plants.

Seismic fragility analysis of steel reinforced concrete frame‐bent structures in CAP1400 nuclear power plant

SummaryThis paper evaluates the seismic performance of the frame‐bent structure of a conventional island main building in a nuclear power plant (NPP) using the CAP1400 NPP project of Rongcheng Shidaowan in Shandong Province as a case study. A 12‐pin 3‐span prototype structural model of a steel reinforced concrete (SRC) frame‐bent main building is established in ABAQUS, and its seismic fragility is analyzed. The results show that under the conditions of an 8‐degree frequent earthquake and an 8‐degree fortification earthquake, the transcendental probability under the four limit states of the structural seismic fragility matrix with Sa as intensity measure (IM) is higher than that with peak ground acceleration (PGA) as IM. Moreover, under an 8‐degree rare earthquake, the structure has completely exceeded the normal service performance level and has reached the service performance level of temporary use and repair. The seismic damage index of the structure demonstrates that the structure's seismic damage does not exceed the medium damage state under a small earthquake. However, under the action of a moderate earthquake, the seismic damage of the structure ranges from moderate to serious damage. When subjected to large earthquakes, the seismic damage of the structure is basically in a state of serious damage.

Identification method of internal leakage in nuclear power plants valves using convolutional block attention module

Valves are indispensable fluid control devices widely employed in Nuclear Power Plant (NPP) structures. Due to prolonged exposure to high-temperature environments, the internal sealing components become susceptible to thermal deformation or wear, leading to compromised sealing and potential leakage accidents in the valves. Therefore, detecting internal leakage in valves holds great significance. We propose an effective acoustic emission detection device designed for identifying internal leakage in valves within high-temperature environments. A valve internal leakage monitoring system based on the Acoustic Emission (AE) method is developed, utilizing high-temperature piezoelectric transducers as AE sensors. To efficiently and rapidly identify leakage states, an artificial intelligence (AI) algorithm incorporating a Convolutional Neural Network (CNN) with the Convolutional Block Attention Module (CBAM) is employed. The approach integrates both a channel attention module and a spatial attention module to capture global and local features, respectively. Normalized spectral data is used as the training set to optimize the CNN parameters. The performance of this method is then compared with BAM and SENET. The results indicate that CBAM can effectively enhance and suppress features by allocating weights among channels according to task requirements. Moreover, it can more efficiently identify multiple damage features under the constraints of limited computing resources.

The European GO-VIKING project on flow-induced vibrations: overview and current status

Abstract The interaction between cooling fluid and solid structures (rods, tubes) in nuclear power plants may lead to flow-induced vibrations (FIV), causing material fatigue, fretting wear, and eventually loss of component integrity. This can cause further safety issues as well as substantial standstill costs due to longer or unplanned outages. With the growing computational power, the application of modern 3D numerical simulation tools for the accurate prediction of FIV phenomena is rapidly increasing. In 2022, the GO-VIKING (Gathering expertise On Vibration ImpaKt In Nuclear power Generation) project received a grant within the Horizon Europe research and innovation funding program. Sixteen European and two US partners started their collaboration in the field of FIV experiments and analysis. Over four years, the GO-VIKING project investigates FIV phenomena occurring in nuclear reactor cores and steam generators under single- and two-phase flow conditions. The project’s main objectives are to expand the expertise in the field of FIV through generation of new experimental and high-resolution numerical data; development, improvement, and validation of fluid-structure interaction (FSI) methods for FIV evaluation; training stakeholders in the application of these methods; and synthesizing guidelines for the prediction and assessment of FIV phenomena in nuclear reactors. This paper provides an overview of the GO-VIKING objectives, scientific program, as well as of the main scientific achievements in the first project year.

Copula-based cloud analysis for seismic fragility and its application to nuclear power plant structures

This paper presents a copula-based cloud analysis method for seismic fragility assessment with practical applications for nuclear power plant structures. Copula theory enables the flexible and efficient modeling of joint probability density function (PDF) for critical parameters, specifically the intensity measures (IM) and the engineering demand parameters (EDP). It decomposes the joint PDF into the product of marginal PDFs and copula density function, which are then separately fitted to the data. The study showcases the versatility of parameterized joint distribution models in estimating optimal distribution parameters derived from structural seismic analysis data. By utilizing copula functions, the method calculates conditional EDP distributions given IM, facilitating the construction of seismic fragility curves. The research compares various copula models’ suitability for fitting fragility curves, employing methods like Akaike information criterion (AIC) for model selection. The study provides a comprehensive framework for seismic fragility assessment, from random sampling of structural parameters and seismic data to nonlinear finite element modeling and copula-based parametric cloud analysis. The application to prestressed concrete containment vessel (PCCV) structures exemplifies the method’s adaptability. This innovative approach not only highlights the flexibility of copula-based analysis in assessing structural performance under different dynamic scenarios but also offers invaluable insights for performance-based earthquake engineering (PBEE) and safety assessments.

Investigations on seismic performance of nuclear power plants equipped with an optimal BIS-TMDI considering FSI effects

This paper introduces a base isolation system-tuned mass damper inerter (BIS-TMDI) hybrid system to the AP1000 nuclear power plant (NPP), which reduces seismic damage potential of the NPP structure. The effects of fluid-structure interaction (FSI) caused by the passive containment cooling system water storage tank (PCCWST) on NPP's seismic performance are investigated. The FSI of water tank theoretical model is considered based on the Housner's model, and a series of time history analyses are performed to prove the rationality of the proposed model. Three single-objective optimization strategies are employed to minimize the relative displacement variance and absolute acceleration variance of the upper structure, as well as the filtered energy index (FEI). Furthermore, a multi-objective optimization strategy considering all these three indexes is proposed to obtain optimal parameters of vibration control. The influence of vibration control strategies on the relative deformation and acceleration of the upper structure is explored with various water level ratios. The analytical results indicate that the proposed BIS-TMDI strategy has significantly reduced the NPP structure's seismic response. The effectiveness of the vibration control strategy is influenced by the water level ratio, emphasizing the significance of designing an appropriate water level ratio to reduce NPP structure's seismic response.

Seismic fragility analysis of nuclear containment structure using Bayesian logistic regression model

Nuclear containment structure serves as the critical shielding structure in nuclear power plants, and must maintain structural integrity under various conditions. Following the Fukushima nuclear accident, seismic safety of nuclear power plant structures has become a significant concern in nuclear engineering community. This study presents seismic fragility analysis of the nuclear containment structure subjected to far-fault ground motion based on Bayesian logistic regression model. A refined finite element model of the nuclear containment structure was developed using layered shell element with considering material nonlinear behavior. The maximum tensile strain and roof drift of the nuclear containment structure were analyzed. MATLAB scripts for the seismic fragility analysis method based on Bayesian logistic regression model were developed. Posterior samples for the parameter of fragility function parameters were obtained using developed scripts. The sensitivity of the proposed Bayesian logistic regression based seismic fragility analysis method was verified. Seismic fragility curves of the nuclear containment structure obtained with the minimized sum of squared error, maximum likelihood estimation and Bayesian logistic regression methods were compared in depth.

УСТОЙЧИВОСТЬ ЗДАНИЙ И СООРУЖЕНИЙ ЗАПОРОЖСКОЙ АТОМНОЙ ЭЛЕКТРОСТАНЦИИ ПРИ ИСПОЛЬЗОВАНИИ АРТЕЗИАНСКОЙ ВОДЫ ДЛЯ ТЕХНОЛОГИЧЕСКОГО ПРОЦЕССА СТАНЦИИ

The article provides an assessment of ensuring safe operation, water supply, as well as the safety of personnel and the environment during the operation of the Zaporizhia NPP. It has been established that the exploitation of groundwater for the technical water supply of the NPP and its infrastructure facilities, both under constant and periodic drainage regimes, will lead to the formation of depression funnels, the size of which will depend on the magnitude (flow rate) of the water intake, the layout of water intake facilities and specific hydrogeological conditions of the groundwater deposit. It is shown that it is necessary to ensure not only the strength of the soils under the structures of buildings and structures of nuclear power plants, but also the reliable functioning of all systems that control the nuclear process. The necessity of monitoring the stability of geological and geophysical conditions in the area and at the sites of nuclear power plants in the monitoring mode has been determined. The monitoring is comprehensive. It includes geotechnical, surveying and geodetic, hydrogeological, and seismological observations. The forecast of the development of a depression funnel in time and space should be carried out by analytical methods or modeling, taking into account changes in the conditions of the relationship of aquifers with each other and environmental components. It has been revealed that extending the further operation of the NPP is an expensive and unsafe process. It is not possible to upgrade old power units to modern ones in terms of safety in nuclear energy. The power units of the Zaporizhia NPP are not ready for further operation, therefore, it is necessary to start developing a project for decommissioning power units No. 1-4 of the NPP.The research results can be useful in predicting the possible development of karst-suffusion processes under the influence of intensive groundwater drainage, organizing monitoring of the possible development of these processes, including them in the integrated monitoring of this facility. It will make it possible to justify the scale and regime of rational water extraction, taking into account the requirements for environmental protection and the specifics of further operation of the NPP.

A Study on the Statistical Analysis and Artificial Intelligence Model to Improve the Reliability of Safety Inspection

In this study, safety diagnosis results from underground and nuclear power plant structures were collected to evaluate the reliability of on-site structural safety assessments. The analysis of these results revealed a discrepancy between the compressive strength determined by rebound hardness and the core compressive strength, with the former typically being evaluated higher than the latter. Additionally, existing strength prediction models did not adequately explain field data, whereas artificial intelligence models, particularly the support vector machine model, demonstrated improved accuracy and reduced error rates. This indicated the superior performance of support vector machine models in this context.

Methods of assessing the seismic resistance of building structures and nuclear power plant structures in Lira-SAPR program

This article proposes a methodology for assessing seismic resistance and determining the reserve of seismic resistance, taking into account the interaction of the structure with the foundation for buildings and structures of power units, strength and deformation criterion of workability, etc. During verification calculation, three criteria of strength of workability, compiled according to the General principles of ensuring reliability and structural safety of buildings and structures, were considered. The seismic calculation was performed according to the linear-spectral theory of seismic resistance. In LIRA-SAPR program, considerationoftheinfluenceofthebaseisimplementedaccordingtoBuildingnorm DBN B.1.1-12-2014. Building Construction in seismic areas of Ukraine.
 The method of calculating the value of the ultimate seismic resistance of HCLPF and the method of constructing damage curves are proposed. To determine the integral parameter HCLPF, which characterizes the level of seismic resistance of this element, a calculation analysis of the seismic resistance of elements of operating nuclear power plants was carried out within the framework of the method of ultimate seismic resistance. The methodology and verification examples for determining the values of FS and HCLPF parameters for reinforced concrete structuresin LIRA-SAPR programare proposed.

Shaking Table Testing of a Scaled Nuclear Power Plant Structure with Base Isolation

To investigate the seismic performance and isolation effect of a high-temperature gas-cooled reactor, a 1/20 scale model including a reactor, a spent-fuel plant, and a nuclear auxiliary plant was fabricated. In addition, 220 mm lead-rubber bearings were designed and produced for use in the shaking table test, which included both isolated and nonisolated conditions. Two historical earthquake records and three artificial earthquake motions were used to input the ground motion in the tests. The results demonstrated that the seismic performance of the plant was better and that the structure was in an elastic state, under a safe shutdown earthquake event. Isolation bearings were found to effectively reduce the dominate frequency of the structure. The acceleration amplification factor of the superstructure was found to be less than 1. The isolation test results showed that the peak of the floor response spectrum at the pressure vessel support was less than 0.1 g. In the nonisolation test, the peak of the floor response spectrum was greater than 1 g. In the isolation test, the relative displacement of the structure was less than 1.1 mm, which was relatively small. The structure maintained a good isolation performance and exhibited improved safety under extreme ground motion.

Benchmarking Standard and Micromechanical Models for Creep and Shrinkage of Concrete Relevant for Nuclear Power Plants.

The creep and shrinkage of concrete play important roles for many nuclear power plant (NPP) and engineering structures. This paper benchmarks the standard and micromechanical models using a revamped and appended Northwestern University database of laboratory creep and shrinkage data with 4663 data sets. The benchmarking takes into account relevant concretes and conditions for NPPs using 781 plausible data sets and 1417 problematic data sets, which cover together 47% of the experimental data sets in the database. The B3, B4, and EC2 models were compared using the coefficient of variation of error (CoV) adjusted for the same significance for short-term and long-term measurements. The B4 model shows the lowest variations for autogenous shrinkage and basic and total creep, while the EC2 model performs slightly better for drying and total shrinkage. In addition, confidence levels at 5, 10, 90, and 95% are quantified in every decade. Two micromechanical models, Vi(CA)2T and SCK CEN, use continuum micromechanics for the mean field homogenization and thermodynamics of the water-pore structure interaction. Validations are carried out for the 28-day Young's modulus of concrete, basic creep compliance, and drying shrinkage of paste and concrete. The Vi(CA)2T model is the second best model for the 28-day Young's modulus and the basic creep problematic data sets. The SCK CEN micromechanical model provides good prediction for drying shrinkage.

Numerical investigation on seismic performance of reinforced rib-double steel plate concrete combination shear wall

Double steel plate concrete composite shear wall (SCSW) has been widely utilized in nuclear power plants and high-rise structures, and its shear connectors have a substantial impact on the seismic performance of SCSW. Therefore, in this study, the mechanical properties of SCSW with angle stiffening ribs as shear connections were parametrically examined for the reactor containment structure of nuclear power plants. The axial compression ratio of the SCSW, the spacing of the angle stiffening rib arrangement and the thickness of the angle stiffening rib steel plate were selected as the study parameters. Four finite element models were constructed by using the finite element program named ABAQUS to verify the experimental results of our team, and 13 finite element models were established to investigate the selected three parameters. Thus, the shear capacity, deformation capacity, ductility and energy dissipation capacity of SCSW were determined. The research results show that: compared with studs, using stiffened ribs as shear connectors can significantly enhance the mechanical properties of SCSW; When the axial compression ratio is 0.3–0.4, the seismic performance of SCSW can be maximized; with the lowering of stiffener gap, the shear bearing capacity is greatly enhanced, and when the gap is lowered to a specific distance, the shear bearing capacity has no major affect; in addition, increasing the thickness of stiffeners can significantly increase the shear capacity, ductility and energy dissipation capacity of SCSW. With the rise in the thickness of angle stiffening ribs, the improvement rate of each mechanical property index slows down. Finally, the shear bearing capacity calculation formula of SCSW with angle stiffening ribs as shear connectors is derived. The average error between the theoretical calculation formula and the finite element calculation results is 8% demonstrating that the theoretical formula is reliable. This study can provide reference for the design of SCSW.

A Review of Concrete Carbonation and Approaches to Its Research under Irradiation

The current state of knowledge on concrete carbonation has proven that this phenomenon is one of the key factors influencing the reinforced concrete durability reduction during the operational period. To date, the carbonation process has been researched quite deeply; however, the dependence of its course on a variety of external and internal factors poses a significant problem in service life predictions for concrete constructions. The development of nuclear infrastructure around the world in recent years has set scientists the task of investigating such processes in conditions different from those usual for industrial and civil construction. In particular, information in open sources on the course of the carbonation process under irradiation conditions is insufficient. The manuscript analyzes the existing data on concrete carbonation, including a review of the main methods for studying the carbonation process, key factors influencing the course of this process, applied methods of mathematical analysis, predictive models of service life, dynamics of carbonation development, and the application of such analytical models in practice. The available information about the carbonation process under various types of irradiations on the causes, dynamics, and mechanisms of carbonation and corrosion processes occurring in reinforced concrete during operation is also considered. Based on the results of the analysis carried out in the study, recommendations are given for further development in the research field of carbonation process in concrete structures of nuclear power plants in order to comprehensively predict their service life.