Nuclear Power Plant Piping Systems 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.

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.

A multi-state Semi-Markov model for nuclear power plants piping systems subject to fatigue damage and random shocks under dynamic environments

Fatigue degradation is a vital damage mechanism for piping systems in nuclear power plants (NPPs). The accurate fatigue reliability evaluation of piping systems is significant for assessing the safety and reliability of an NPP. In this work, we presented a multi-state model for the fatigue degradation process of piping systems in NPPs, which accounts for the effects of random shocks and dynamic environments. The fatigue degradation of the piping was described by a Semi-Markov process, which allows accounting for generic distributions of the holding times of the system states. The arrival of the shocks was assumed to be governed by a Poisson process. The dynamic environment in which the piping is located was assumed to be governed by a Markov process. The analytical solution of the time-dependent state probability of the piping was derived. We developed a Monte Carlo (MC) algorithm for the simulation of the stochastic process describing the integrated stochastic evolution of the dynamic environment, system degradation and random shocks, which was applied to verify the correctness of the proposed model. The effectiveness of the proposed model was demonstrated by applying the model to numerical illustrations.

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.

Improvement on optimal design of dynamic absorber for enhancing seismic performance of nuclear piping using adaptive Kriging method

For improving the seismic performance of the nuclear power plant (NPP) piping system, attempts have been made to apply a dynamic absorber (DA). However, the current piping DA design method is limited because it cannot provide the globally optimum values for the target design seismic loading. Therefore, this study proposes a seismic time history analysis-based DA optimal design method for piping. To this end, the Kriging approach is introduced to reduce the numerical cost required for seismic time history analyses. The appropriate design of the experiment method is used to increase the efficiency in securing response data. A gradient-based method is used to efficiently deal with the multi-dimensional unconstrained optimization problem of the DA optimal design. As a result, the proposed method showed an excellent response reduction effect in several responses compared to other optimal design methods. The proposed method showed that the average response reduction rate was about 9% less at the maximum acceleration, about 5% less at the maximum value of the response spectrum, about 9% less at the maximum relative displacement, and about 4% less at the maximum combined stress compared to existing optimal design methods. Therefore, the proposed method enables an effective optimal DA design method for mitigating seismic response in NPP piping in the future.

Numerical investigation on similarity of isothermal and thermal flow mixing in a horizontal T-junction configuration

Turbulent and stratified mixing flows can cause thermal fatigue in nuclear power plant piping systems. An isothermal Mixed-Fluid-Interaction (MFI) test rig is used to investigate mixing phenomena related to High Cycle Thermal Fatigue in nuclear power plant piping systems. Here, the cold heavy branch pipe fluid of the thermal mixing Fluid Structure-Interaction (FSI) facility is modeled by a sugar or salt solution. Due to limited optical accessibility of the FSI facility a numerical similarity comparison of the flow phenomena occurring in both experimental setups (MFI/FSI) is essential. Thus, Large-Eddy Simulations are carried out which are experimentally validated by applying the Wire-Mesh-Sensor measurement technique (MFI) and temperature measurements (FSI). Numerical LES investigation on similarity between two isothermal cases, three thermal mixing cases, and one geometrically similar scaled-up thermal mixing case is performed when the momentum ratio MR and the Richardson number Ri is kept constant. A very good agreement in the spatial distribution of the time averaged mean and rms values as well as statistical fluctuations is observed between the isothermal and thermal cases. Non-dimensional power spectrum density (PSD) of the mixing scalar fluctuation showed an almost unique profile, indicating the same origin for the occurring periodical fluctuation in all cases. The non-dimensional magnitude of the dominant frequency of the mixing scalar fluctuations in terms of the Strouhal number is in the range of 0.38<Sr<0.40. In the context of similarity, it is shown that the momentum ratio and the Richardson number are mainly responsible for the jet-formation and thus for the down- and upstream flow behaviour. The influence of the inlet Reynolds numbers Rem and Reb as well as the density ratio ΓR on the normalized flow characteristics seem to be negligible in comparison.

Seismic performance limit of nuclear power plant piping system with seismic isolation system considering measurement points: Low-cycle fatigue life

This paper presents the first part of a two-part investigation on the seismic performance limits of nuclear power plant piping systems with seismic isolation systems considering the measurement points and proposes a scheme for low-cycle fatigue life. This part of the study focuses on pipe elbows, which are components vulnerable to seismic loads in crossover piping systems connecting structures with and without seismic isolation systems in nuclear power plants. The low-cycle fatigue life, which is the seismic performance limit of the pipe elbow, was determined by considering the cyclic loading of the low-frequency characteristics caused by the seismic isolation system. Three-inch test specimens of SCH40, which are typical standards for nuclear power plant piping systems, are fabricated herein consisting of a pipe elbow and straight pipes. In the test specimens, the pipe elbow where the load is concentrated and the lengths of the straight pipes that are connected to the elbow may differ from those in field conditions. Therefore, their effects on the seismic performance limit were analyzed using the relationship between the moment and deformation angle. Cyclic loadings were performed at varying amounts of a constant amplitude, and an imaging system was used to measure the moments and deformation angles. The deformed shapes of the test specimens caused by external forces were compared for each measurement point, and the leakage lines and low-cycle fatigue curves were presented using the relationship between the moment and relative deformation angle.

Seismic performance limit of nuclear power plant piping system with seismic isolation system considering measurement points: Damage index

This paper presents the second part of a two-part investigation on the seismic performance limits of nuclear power plant piping systems with seismic isolation systems considering the measurement points and aims to quantitatively express these limits using a damage index based on the dissipated energy. This part of the study focuses on a method to quantitatively express the seismic performance limits needed for seismic fragility evaluations of crossover piping systems connecting structures with and without seismic isolation systems in nuclear power plants. Test specimens for this study were manufactured by welding straight pipes to both ends of a 90° long pipe elbow. The dissipated energy was calculated using the relationship between the moment measured at the center of the pipe elbow and relative deformation angle at each of the straight pipe’s measurement points using an imaging system; this was compared to the dissipated energy calculated from the relationship between force and displacement, and a suitable length of the straight pipe for low-cycle fatigue tests was found. Further, a method of applying a weighting factor to the yield points, which are needed to calculate the damage index, are examined; thus, it was confirmed that the damage index can be used to quantitatively express the seismic performance limits of thin-walled steel pipe elbows.

Fatigue behaviors of the steel pipe tee in a nuclear power plant piping system under in-plane cyclic loading

The safety of a nuclear power plant against earthquakes can be improved by installing seismic isolation devices. As such seismic isolation devices handle seismic loads, however, displacements larger than those generated before the application of such devices may occur. Therefore, the seismic risks are likely to increase for some facilities. In particular, pipes that connect structures with seismic isolation devices to ordinary structures are likely to have higher seismic risks. The piping is affected by the behavior of the two supports, and damage may occur due to the displacement-dominant repeated behavior. In addition, to assess the low-cycle fatigue behavior of the piping system under seismic loads, it is essential to identify the relationships among the number of cycles, moment, and deformation angle until leakage occurs, through an in-plane cyclic loading test. In this study, an in-plane cyclic loading test was conducted to assess the fatigue behavior of a steel pipe tee against seismic loads from the correlation between the moment and the deformation angle.

Anti-seismic characteristics of safety valve and its spring failure prediction analysis

Safety valve is an important guarantee for nuclear power plant system. Its working environment is harsh, and medium is high-temperature corrosive. It is important to study the dynamic characteristics of nuclear safety valve under unsteady condition to improve the stability and safety of nuclear power plant piping system. Using the finite element method, mass simplification and Rayleigh method, the frequency response analysis of the safety valve and the spring is predicted to predict the seismic capacity under the earthquake load. Based on the verification of the dynamic characteristics of the valve body, the stress and strain analysis of the spring is carried out. To explore the failure condition of the safety valve spring, the lateral deviation linearity of the spring was tested and analyzed by setting up the offset test bench. The results show that the safety valve with the above analysis method can meet the working requirements under the shaking condition. The research results provide an important theoretical basis for the design and analysis of nuclear safety valves.

Large-Eddy Simulation of turbulent thermal flow mixing in a vertical T-Junction configuration

Turbulent mixing of warm and cold water streams within a vertical T-junction configuration of a nuclear power plant piping system causes near-wall temperature fluctuations which may lead to High-Cycle Thermal Fatigue (HCTF) failure of the wall material. To predict frequency and amplitude of such temperature fluctuations accurately the method of Large-Eddy Simulation (LES) in the numerical simulation code OpenFOAM is used. As an experimental test case, the turbulent mixing experiment of Kamide et al. [1], where a vertical branch pipe with a cold water stream is connected from below to a horizontal main pipe with a warm water stream to form a vertical T-junction configuration. From the various observed flow patterns, the temperature and velocity data of the ‘wall jet’ are chosen as a reference because it is known to have the highest potential for thermal fatigue. Incomplete mixing and an instability associated with large turbulent structures near the junction region lead to significant temperature fluctuations close to the wall before a vertical thermal stratification further downstream stabilizes the flow. The highest near-wall fluctuation amplitudes are 26% of the temperature difference between the streams. A spectral peak occurs at a Strouhal number of 0.2. The results of the simulations demonstrate, that the mean velocity and temperature profiles of the experiments are well captured, whereas the root-mean-square (RMS) temperature fluctuations deviate from the measurements in some cases, in particular when coarse grids are used. In order to find a lower limit for the required spatial resolution three different numerical meshes with a variable number of cells up to 28 × 106 are investigated. The results demonstrate that with a sufficient mesh resolution the velocity and temperature distributions as well as the spectral peak can be simulated with good agreement to the experimental data.

Enhancement of seismic resilience of piping systems in nuclear power plants using steel coil damper

After the accident at Fukushima, it has become more a mandatory to enhance the safety level of nuclear power plants (NPPs). It has been suggested that the piping systems in NPPs should be improved as part of measures to mitigate severe accidents that may occur when the earthquake ground input motion exceeds the level of the design basis earthquake (DBE). In this study, a steel coil damper (SCD) has been proposed to reduce the vibration of the piping system and ensure the safety of the NPP when a seismic event exceeds the DBE. This study developed an analytical model and design method for the SCD. The basic mechanical characteristics of the SCD were verified through experiments on the specimens designed and manufactured according to design equations developed in this study. The proposed analytical model of SCD is compared with the mechanical characteristics obtained from the experiments. Test results agree well with the analytical mechanical characteristic of the SCD. The actual seismic margin of the piping system with SCDs was evaluated by the analyses. The study results prove that the proposed damper is effective at maintaining the function and ensuring the safety of the NPP under extreme conditions.

Low-cycle fatigue behaviors of the elbow in a nuclear power plant piping system using the moment and deformation angle

Abstract Maintaining the structural integrity of the major devices of a nuclear power plant is perceived as a particularly important issue in relation to the stability of the structure, and the structural integrity of the piping system is especially important for the safety of the nuclear power plant. In this study, the limit state of a steel pipe elbow, which is a fragile part of the nuclear power plant piping system, was defined as leakage, and an in-plane cyclic loading test was conducted. As it is difficult to measure the moment and deformation angle of the steel pipe elbow using conventional sensors, an image measurement system was used. A deformation angle measurement algorithm that uses an image measurement system applied an improved deformation angle measurement algorithm that uses image filter processing to facilitate noise reduction in images as well as the boundary line of the piping system. This study proposed a leakage line and low-cycle fatigue curves using the relationship between the moment and the deformation angle of a 3-in steel pipe elbow.

Seismic fragility evaluation of the base-isolated nuclear power plant piping system using the failure criterion based on stress-strain

Abstract In the design criterion for the nuclear power plant piping system, the limit state of the piping against an earthquake is assumed to be plastic collapse. The failure of a common piping system, however, means the leakage caused by the cracks. Therefore, for the seismic fragility analysis of a nuclear power plant, a method capable of quantitatively expressing the failure of an actual piping system is required. In this study, it was conducted to propose a quantitative failure criterion for piping system, which is required for the seismic fragility analysis of nuclear power plants against critical accidents. The in-plane cyclic loading test was conducted to propose a quantitative failure criterion for steel pipe elbows in the nuclear power plant piping system. Nonlinear analysis was conducted using a finite element model, and the results were compared with the test results to verify the effectiveness of the finite element model. The collapse load point derived from the experiment and analysis results and the damage index based on the stress-strain relationship were defined as failure criteria, and seismic fragility analysis was conducted for the piping system of the BNL (Brookhaven National Laboratory) - NRC (Nuclear Regulatory Commission) benchmark model.

Vibration Control of Nuclear Power Plant Piping System Using Stockbridge Damper under Earthquakes

Generally the piping system of a nuclear power plant (NPP) has to be designed for normal loads such as dead weight, internal pressure, temperature, and accidental loads such as earthquake. In the proposed paper, effect of Stockbridge damper to mitigate the response of piping system of NPP subjected to earthquake is studied. Finite element analysis of piping system with and without Stockbridge damper using commercial software SAP2000 is performed. Vertical and horizontal components of earthquakes such as El Centro, California, and Northridge are used in the piping analysis. A sine sweep wave is also used to investigate the control effects on the piping system under wide frequency range. It is found that the proposed Stockbridge damper can reduce the seismic response of piping system subjected to earthquake loading.

An Itô model of pit corrosion in pipelines for evaluating risk-informed decision making

The purpose of this paper is to present a stochastic model based on the Itô differential equation to the time evolution of pit corrosion depth in nuclear power plant piping systems. The stochastic solution obtained by analytical and numerical methods (for comparison purposes) is compared to the deterministic solution, which is obtained by setting the stochastic term of the Itô equation equal to zero. The study of extended qualified useful life of nuclear power plants has two components, one deterministic and another probabilistic and one complements the other. The evaluation based on deterministic methods defines the difference between the current state and condition of the item in the qualifying phase, but does not define its probability of continuing performing its function adequately for a period longer than the one defined by its qualified life. The corrosion process is modeled by the Itô stochastic equation, which describes the time evolution of pit corrosion depth among different states. The Itô equation is determined by the knowledge of two functions known as the drift and diffusion coefficients. This work shows that these functions can be used to estimate pit corrosion depths from plant inspection data. An Itô modeling of pit corrosion can also be developed for any component of a nuclear power plant for which corrosion depth data is available. The above proposed approach provides a precise and easy way in which pit corrosion depth can be monitored over time, which is critical for developing reliability and risk-informed models for inspection and maintenance planning of corroded pipelines. The stochastic model presented is developed considering the influence of the Gaussian white noise on corrosion. The equations of the stochastic model are solved analytically based on the Itô integral. The maximum operating time of a piping before its wall reaches the minimum allowable thickness is calculated. This study uses pit corrosion depth data from the Angra I nuclear power plant.

Experimental fracture studies on carbon steel elbows with and without internal pressure

Pipe bends or elbows are commonly used components for nuclear power plant piping system. In service, these piping components are subjected to internal pressure in addition to bending loads and the internal pressure is known to have a significant effect on the load carrying capacity of these components. Hence, a systematic study was carried out to investigate and quantify the effect of internal pressure on the fracture behaviour of elbows used in nuclear power plant piping system. Fracture studies were conducted on five 219 mm diameter carbon steel elbows with and without internal pressure under in-plane opening moment. The investigations have shown that the presence of a circumferential notch at the intrados has a more detrimental effect on the fracture behaviour of the elbow, when compared with the presence of an axial notch at the intrados. It is also found that internal pressure plays a significant role in reducing the ovalization.

다지점 진동대를 이용한 원자력발전소 배관계통의 내진성능실험

In this study, dynamic characteristics and seismic capacity of the nuclear power plant piping system are evaluated by model test results using multi-platform shake table. The model is 21.2 m long and consists of straight pipes, elbows, and reducers. The stainless steel pipe diameters are 60.3 mm (2 in.) and 88.9 mm (3 in.) and the system was assembled in accordance with ASME code criteria. The dynamic characteristics such as natural frequency, damping and acceleration responses of the piping system were estimated using the measured acceleration, displacement and strain data. The natural frequencies of the specimen were not changed significantly before and after the testing and the failure and leakage of the piping system was not observed until the final excitation. The damping ratio was estimated in the range of 3.13 ~ 4.98 % and it is found that the allowable stress(345 MPa) according to ASME criteria is 2.5 times larger than the measured maximum stress (138 MPa) of the piping system even under the maximum excitation level of this test.

국부 감육과 균열이 발생한 TP316 스테인리스강 배관의 파괴거동에 관한 실험적 연구

Although nuclear power plant piping system is designed conforming to design specifications, the piping systems are deteriorated with increase in service life. In this study, monotonic and cyclic loading tests were carried out on TP316 stainless steel pipe specimens, and the effect of local wall thinning and cracking on failure behavior was investigated. In the tests, 0%, 35% and 75% wall thinning and cracking of initial thickness were artificially introduced to inside elbow and straight pipe specimens, and internal pressures of 20MPa were applied to simulate real operation condition. From the test results, the effect of local wall thinning and cracking on failure mode, ultimate load, number of cycle and strain energy was presented, and maximum bending moment was compared with allowable bending moment calculated by ASME code.