Nuclear Power Plant Piping Research Articles

Failure criterion and seismic fragility evaluation of isolated nuclear power plant piping system using BP neural network

The pipeline connecting the isolated nuclear island and the non-isolated turbine building is prone to significant relative displacements during intensive earthquakes. Generally, elbows closest to the isolation layer are considered to be at particularly high risk of seismic damage. The failure criterion is the critical part in determining the damage state of elbows throughout the loading cycle. Therefore, in this study, a three-tier, four-category failure criterion was firstly proposed in response to the damage patterns of elbows. A BP neural network model was subsequently developed using extensive simulation data derived from a verified numerical model, accompanied by influence analysis of the three failure indices. Finally, a fragility evaluation was conducted based on the seismic response of the piping system, and a comparison is made between the new criterion and the existing ones. The results indicate that the new failure criterion demonstrates comprehensiveness and safety, enabling a more detailed division of damage before elbow failure and providing more conservative estimates upon failure occurrence. Additionally, the study finds that under the current design of nuclear power plants, elbows across isolation layers fail to meet the requirements for safe shutdown earthquakes in certain directions. By adjusting the elbow parameters, the resistance of the elbows to damage under seismic conditions can be improved to some extent.

Engineering estimate of plastic collapse loads for cracked pipe bends under in-plane and out-of-plane bending

Nuclear power plant piping systems frequently encounter complex loads due to factors like internal pressure, self-weight, structural supports, and operational conditions. Within these systems, pipe bends play a crucial role in releasing energy through deformation. Cracks pose a risk to the integrity of the system, necessitating the calculation of the plastic collapse load for cracked pipe bends. This study presents estimations for the plastic collapse load of cracked pipe bends under pure out-of-plane bending moments and simultaneous in-plane and out-of-plane bending moments. Cracks located in the intrados, extrados, and crown of pipe bends were considered. The estimations for plastic collapse load are proposed based on the observed similarity between plastic collapse load and plastic deformation under both in-plane and out-of-plane bending conditions. For combined in-plane and out-of-plane bending moment conditions, circular and parabolic relationships have been proposed for a straightforward plastic collapse load prediction. Additionally, the estimations for plastic collapse loads were derived from the results of finite element analysis conducted on crack-free pipe bends, demonstrating its applicability to cracked pipe bends as well.

Effects of microalloying element addition on mechanical properties of SA508 Gr.1A low-alloy steels

SA508 Gr.1A low-alloy steel is being considered as a candidate material for main steam line piping in nuclear power plants. Therefore, improving its strength and toughness is essential for enhancing the leak-before-break (LBB) margin. In this study, six types of model alloys were fabricated by varying the contents of microalloying elements (C, Cu, B, Ti, and Nb) to enhance the mechanical properties of the specimens. The addition of a few ppm of B led to the formation of a fine-grained low-temperature transformation microstructure, resulting in the highest strength among the model alloys. However, the addition of Nb and Ti increased the formation of coarse ferrite, significantly decreasing the strength of the alloys. Reducing the C content while adding a small amount of B simultaneously maintained strength and enhanced toughness. Furthermore, the LBB margins of model alloys and commercial steel were evaluated to validate the influence of varying microalloying content. The model alloys exhibited a substantial increase in yield strength and fracture resistance, resulting in a more than 10% increase in the LBB margin. Notably, the LBB margin of the alloy with 15 ppm B was 1.39, approximately 25% higher than that of commercial steels.

Dynamics optimization of small branch pipes in nuclear power plants based on machine learning algorithms

Vibration fatigue failure of small branch pipes poses a great threat to the safe operation of nuclear power plants. However, the transient and wide-band vibration problems are not adequately considered in ASME and RCC-M code, resulting in repeated fatigue failures. In addition, the current research mainly focuses on vibration test methods and fatigue analysis methods, neglecting the study of pipeline vibration characteristics. Therefore, innovative approaches were essential for effectively managing complex dynamic loads. In this study, an innovative approach combining backpropagation artificial neural networks (BP-ANN) and non-dominated sorting Genetic Algorithm II (NSGA-II) was proposed to optimize the vibration of these pipes. The goal was mitigating vibration-induced failures by enhancing operational stability. The methodology progressed through several key stages. Firstly, BP-ANN was utilized for regression analysis, correlating pipe characteristics to vibration effects. Through regression analysis, the complex interrelationships governing the pipes' dynamic behavior was revealed. Subsequently, based on the regression model, NSGA-II was used to derive an optimal combination of design parameters to minimize the vibration response. The proposed technique was validated on an L-shaped cantilevered pipe via finite element simulations and physical experiments. The analysis case shows that the BP-ANN model demonstrated excellent accuracy in predicting vibration responses. Meanwhile, NSGA-II successfully revealed the trade-offs between conflicting objectives, generating a Pareto-optimal set balancing stability under different excitation directions. This study highlights the potential of machine learning methods for dynamic optimization of small branch pipes in nuclear power plants, and verifies the accuracy and effectiveness of the method through experiments. The research results provide a new idea for small branch pipe design and vibration control of nuclear power plant, which contributes to enhancing the safety and reliability of small branch pipes.

Reliability Analysis of PAUT Based on the Round-Robin Test for Pipe Welds with Thermal Fatigue Cracks.

Thermal fatigue cracks occurring in pipes in nuclear power plants pose a high degree of risk. Thermal fatigue cracks are generated when the thermal fatigue load caused by local temperature gradients is repeatedly applied. The flaws are mainly found in welds, owing to the effects of stress concentration caused by the material properties and geometric shapes of welds. Thermal fatigue pipes are classified as targets of risk-informed in-service inspection, for which ultrasonic testing, a volumetric non-destructive testing method, is applied. With the advancement of ultrasonic testing techniques, various studies have been conducted recently to apply the phased array ultrasonic testing (PAUT) method to the inspection of thermal fatigue cracks occurring on pipes. A quantitative reliability analysis of the PAUT method must be performed to apply the PAUT method to on-site thermal fatigue crack inspection. In this study, to evaluate the quantitative reliability of the PAUT method for thermal fatigue cracks, we fabricated crack specimens with the thermal fatigue mechanism applied to the pipe welds. We performed a round-robin test to collect PAUT data and determine the validity of the detection performance (probability of detection; POD) and the error in the sizing accuracy (root-mean-square error; RMSE) evaluation. The analysis results of the POD and sizing performance of the length and depth of thermal fatigue cracks were comparatively evaluated with the acceptance criteria of the American Society of Mechanical Engineers Code to confirm the effectiveness of applying the PAUT method.

Reliability Analysis of Pipe Wall Thinning based on Quantification of Ultrasonic Testing

Piping in nuclear power plants is subject to corrosion and erosion caused by the interaction with fluids it carries. In order to prevent accidents such as pipe rupture, it is necessary to non-destructively inspect the thickness of the pipe wall and predict the remaining pipe life. In contrast, signals obtained by non-destructive inspection are affected by various uncontrollable and unknown factors, which will result in uncertainty in evaluating pipe wall thickness. Ignoring the uncertainty would lead to a large error in the pipe reliability assessment. Therefore, it is important to develop a reasonable pipe wall thinning management that can takes the uncertainty of non-destructive testing into consideration. Based on this background, this study aimed to develop a piping wall thinning prediction model that accounts for the uncertainty in evaluating pipe wall thickness and to evaluate its applicability to ultrasonic testing. At first, ultrasonic tests were performed to measure the thickness of specimens simulating flow accelerated corrosion. The results of the measurements were statistically analysed to obtain a mathematical model correlating the evaluated and true thickness. A numerical model to predict the reduction of wall thinning with its uncertainty is developed. Compared to the conventional method, the proposed method is able to predict wall thickness with high accuracy.

Ultimate Strength of a Thin Wall Tee for Sodium Cooled Fast Reactors Under Seismic Loads

Abstract It is a key issue to clarify failure mode of piping in nuclear power plants and its safety margin included in the design practice. In this study, a series of dynamic failure tests of a thin wall tee for sodium-cooled fast reactors was conducted. Some discussions related to the thin wall tee are provided on the failure mode under seismic load, dynamic response behavior, accuracy of fatigue prediction by dynamic elastic-plastic analysis, and a safety margin of fatigue evaluation which is included in conventional design rule.

Numerical study of thermal mixing phenomenon using partially averaged Navier-Stokes

Thermal mixing between cold and hot streams of fluids in a T-junction may lead to high-cycle thermal fatigue in nuclear power plant piping material. Temperature fluctuations near the wall surface are the main reason for this degradation. Computational Fluid Dynamics (CFD) using space and energy-resolving turbulence modeling paradigms can predict this transient behavior. In this study Partially averaged Navier-Stokes (PANS) turbulence method is used to simulate the Vattenfall T-junction test case. Two different values for the filter control parameter fk = 0.5 and fk = 0.25 are used. Calculations results for velocity and temperature fields are compared with Large Eddy Simulations (LES) and available experimental data. For fk = 0.5, time-averaged and root mean square (RMS) temperatures show lower values than the LES and experimental data. For fk = 0.25 both time-averaged and RMS temperature results are close to the LES and experimental values. The Power Spectral Density (PSD) of the temperature signal shows peaks in the case of both PANS-0.25 and LES in the range of 3 to 4 Hz, which is in line with the experimental results. It can be concluded that the PANS method has the capability to resolve temperature and velocity fields for thermal mixing phenomenon in T-junction pipes and can be used for the assessment of high-cycle thermal fatigue.

Experimental study on seismic responses of tuned mass damper-applied real-scale piping system via shaking table tests

To secure the nuclear power plant (NPP) safety, it is essential to improve the seismic performance of piping systems. Accordingly, to achieve this purpose effectively, applying a tuned mass damper (TMD) to the NPP piping system has been attempted, but this is mostly limited to numerical analysis studies. Therefore, this study aimed to prove the TMD effect of reducing seismic responses of the real-scale NPP piping system through shaking table tests. As a result, it was confirmed that the TMD suppressed target and higher-order mode responses of the piping through the shaking table tests. In particular, the response control effects of the target mode and the higher-order mode were more pronounced when the earthquake motion intensity was larger. The application of the TMD showed reducing relative displacement responses in the harmonic motion excitation by about 80%-90%. This reduced peak relative displacement and normal stress responses by 33%-42% and 25%-37% on average, respectively, under random vibration excitation. Under the seismic excitation, peak relative displacement and normal stress responses were reduced by 7%-14% and 5%-27% on average, respectively. Finally, piping and piping-TMD finite element analysis models were successfully created and validated based on the shaking table test results.

A review on crack growth simulation for estimating failure probability of pipes in nuclear power plants

A review on crack growth simulation for estimating failure probability of pipes in nuclear power plants

Evaluation of seismic margins for nuclear power plant piping systems using ratcheting criterion

In the wake of recent beyond design basis earthquakes, it is very essential to evaluate seismic margins of nuclear power plant piping systems. Better seismic margins will ensure the safety in the event of less probable bigger seismic event beyond consideration of design level. At present, permissible codal stress limits are used to evaluate these margins and it has been experimentally demonstrated that ratcheting is the significant failure mode for piping systems subjected to reversing dynamic load. Hence, it is more prudent to use actual failure criterion for evaluation of seismic margins. In this paper, a detailed framework is provided for evaluation of seismic margins of piping systems using ratcheting criterion. As ratcheting is the potential mode of failure for piping systems under earthquake load, two strain-based performance levels based on code practice and experience gained with experiments carried out by Bhabha Atomic Research Centre, are used for margin assessment. It is usual practice to carry out stress-based response spectrum analysis for design and qualification of piping systems under earthquake load. In order to offer a practical, simpler and feasible solution for ratcheting evaluation, the same response spectrum analysis in iterative way, can be used to evaluate ratcheting in the piping systems. Experimentally validated numerical procedure is used for ratcheting evaluation. This procedure is demonstrated for a piping system of a typical nuclear power plant. Seismic margins for the piping system are evaluated using codal provision and two ratcheting based performance levels for deterministic and probabilistic methods.

Review of Creep-Thermomechanical Fatigue Behavior of Austenitic Stainless Steel

Research on the creep-thermomechanical fatigue (CTMF) behaviors of austenitic stainless steel for nuclear power plant pipelines is reviewed in the present paper. The stress response behavior, the main damage mechanisms, including thermomechanical fatigue, creep, oxidation, and dynamic strain aging (DSA), as well as the effects of strain dwell type, dwell time, and temperature-strain phase angle on fatigue life behavior of austenitic stainless steel under CTMF loading conditions are systematically discussed, and the coupled effects of various damage mechanisms are revealed. It is emphasized that CTMF is closer to the actual service condition of nuclear power plant pipes. It is pointed out that the traditional method of life design based on the isothermal fatigue test data is not conservative. Finally, the research on CTMF behaviors of austenitic stainless steel for nuclear power plant is summarized and prospected.

The Impact of Changes in the Conditions of Fastening of Steel Supporting Structures of Nuclear Power Plants Equipment and Piping on Their Seismic Resistance

The seismic resistance of nuclear power plants equipment and piping is determined, inter alia, by the seismic resistance of their steel supporting structures. The linear-spectral method, which involves using the results of the modal analysis of the structure under consideration, is widely used to assess the seismic resistance of these supporting structures. During the modal analysis, the structure’s dynamic characteristics are researched (in particular, the modes and values of natural oscillation frequencies). The dynamic characteristics of steel supporting structures affect the number of seismic loads that will be transmitted to them during an earthquake. The value of dynamic characteristics, among other issues, is influenced by the conditions of the steel supporting structures fastening. Therefore, it is relevant to research the impact of changes in the conditions of fastening of steel supporting structures of nuclear power plant equipment and piping on their seismic resistance. The paper gives the results of the research of dynamic characteristics, as well as the stress-strain state of steel supporting structures of nuclear power plant equipment and piping during changes in the conditions of their fastening

Development of digital tools to enable remote ultrasonic inspection of fusion reactor in-vessel components

The feasibility of a new approach for pipe inspection has been explored using digital twins to enhance guided wave inspection. Guided wave inspection is well established in the oil & gas industry to remotely screen long lengths of predominately straight pipeline for corrosion. However, the inspection of complex pipe geometries remains a challenge. Nuclear fusion facilities are one such potential application. Fusion reactors have a network of many kilometres of service pipes with complex features, including multiple pipe bends. Some of these pipes could be used for actively cooling components such as the first wall and divertor. Guided ultrasonic wave inspection has the significant advantage of offering 100% coverage of the pipe wall over tens of metres of pipe from a remote test location. This is a highly attractive feature, particularly in the nuclear industry where it is important that human presence in high-risk areas is prohibited due to high radiation doses and temperatures. In this work, finite element wave propagation models have been investigated as digital twins of fusion reactor components. The models have been used to calculate bespoke excitation signals that will allow for full volumetric inspections of these complex pipes to be carried out from a remote location. For the first time, a digital twin technique has been developed that is predicted to successfully correct the distortion in guided wave signals caused by multiple pipe bends. The technique is predicted to yield an order of magnitude improvement in detection capability over conventional guided wave inspection. The digital twin technique presented here therefore shows significant promise for the future inspection of nuclear fusion power plant pipes.

Failure criteria evaluation of steel pipe elbows in nuclear power plant piping systems using cumulative damage models

A steel pipe elbow is an important component in the piping system of a nuclear power plant. Accordingly, its structural performance requires a piping system with high-structural integrity, especially when it is subjected to seismic loads. This component fails when subjected to cyclic loads, such as earthquakes, because of low-cycle fatigue in conjunction with ratcheting. Therefore, steel pipe elbows in piping systems should exhibit sufficient strength, ductility, and energy dissipation capacity characteristics to withstand seismic loads. In this study, various cumulative damage models were used to express quantitatively the failure criteria of the actual failure mode of an SCH40 steel pipe elbow. Experiments were performed until the point of leakage via fatigue cracking, which is the actual failure mode of steel pipe elbows, and in-plane cyclic bending was performed by using displacement control. Based on the experimental results, damage indices were calculated by using various cumulative damage models that considered ductility, energy dissipation, and their combination. Consequently, it was confirmed that the Banon index can express quantitatively the failure criteria for the leakage caused by fatigue cracking.

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.

The Specifics of the Compilation of the Calculated Load Combinations in the Assessment of Seismic Resistance of Steel Supporting Structures of Nuclear Power Plant Equipment and Piping

The seismic resistance of nuclear power plant equipment and piping is determined, inter alia, by the seismic resistance of their steel supporting structures. During the operation of the nuclear power plant power unit, mechanical loads from the elements installed on them are transferred to the steel supporting structures of the equipment and piping. During an earthquake, seismic loads are added to these loads. By state building codes, when considering steel structures in special operating conditions (in particular, exposed to seismic hazards), it is necessary to comply with additional requirements that reflect the features of these structures. Given this, the issue of developing approaches to the compilation of load combinations in assessing the seismic resistance of steel supporting structures of nuclear power plants equipment and piping is acute, taking into account the specific conditions of their operation. The paper is also relevant as it is one of the priority areas of science and technology under the legislation of Ukraine. The development of approaches to the compilation of the calculated load combinations will contribute to the improvement and development of methods for assessing the safety of nuclear power facilities. The paper presents the results of the review of the provisions of state building codes on the calculated combinations of loads when assessing the strength of steel structures. Approaches to the compilation of the calculated load combinations in assessing seismic resistance of steel supporting structures power units of nuclear power plants equipment and piping taking into account the specific conditions of their operation have been developed.

Research of the Stress-Strain State for Steel Support Structures of Nuclear Power Plant Components under Seismic Loads

Seismic resistance of equipment and piping of nuclear power plant units is determined, particularly, by seismic resistance of their steel support structures. A significant number of these structures are located in the reactor building of nuclear power plants and belong to seismic resistance categories I and II. Therefore, such support structures of nuclear power plant equipment and piping, in general, shall perform their functions under seismic hazards that correspond to the safe shutdown earthquake and design-basis earthquakes, respectively.
 Steel support structures of serial equipment (piping valves, electric and pneumatic drives of piping and pump valves, expansion tanks, etc.) are often designed “on site” in the limited conditions of their actual location. Therefore, it is necessary to design steel support structures of a number of equipment taking into account the fact that it can be connected, in particular, to existing piping with their actual routing. In connection with the above, the research of operation for such support structures under seismic loads is relevant.
 The article presents the research results for the spectrum of natural oscillation frequencies of the equipment support structure, and the stress-strain state of the considered structure under seismic loads is determined. At the same time, the parameters during normal operation and maximum design basis accident at the nuclear power plant power unit are considered. The calculated combination of loads, which simultaneously includes two episodic impacts (maximum design basis accident and an earthquake) is considered.

Research on deterioration evolution trend of primary loop piping in nuclear power plant based on fusion health index

The perennial operation of nuclear power primary loop piping leads to deterioration of the piping and the potential risk of leakage. At present, there is no reliable technology to detect the leakage directly. This paper proposes a method to obtain the piping deterioration evolution trend based on the analysis of sets of data that can reflect real state of the piping. The specific method is to quantitatively construct a fusion health index model based on the improved Mahalanobis distance, which integrates the analytic hierarchy process and the entropy weight method. A prediction model combining convolutional neural networks and long short-term memory neural networks is established to make the trend analysis and prediction. The experimental results show that the method can better reflect the actual health state of the piping and effectively predict the deterioration evolution trend, which provides a specific reference value for ensuring the safe and stable operation of equipment.

Effects of the Alloy Design and Fabrication Process on Mechanical Properties of Mo + V-Added SA508 Gr.1A Steel for Main Steam Line Piping in Nuclear Power Plants

Effects of the Alloy Design and Fabrication Process on Mechanical Properties of Mo + V-Added SA508 Gr.1A Steel for Main Steam Line Piping in Nuclear Power Plants