Power Plant Piping Systems Research Articles

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.

Study on Piping Seismic Response Under Multiple Excitation

The uniform response spectrum (URS) analysis method is generally used to seismic qualification of piping systems in nuclear power plants (NPPs). This method can reasonably not only compute dynamic responses of the piping systems exposed to different seismic motions from the building but also tend to overestimate the responses due to an assumption while conservatively considering a variety of the characteristic among input loadings. Increases of design seismic motions for NPPs in Japan by 2–3 times due to the 2011 off Pacific Coast of Tohoku Earthquake resulted that structures, systems, and components such as piping systems are subject to numbers of additional supporting structures to meet the design code requirements. Use of multiple-input analysis methods (independent support motion methods) is expected to bring high degree of precision in dynamic responses than the URS method. However, existence of a handful of experimental tests prevents from utilizing the method in the design process for NPPs in Japan. In order to practically utilize the multiple-input analysis methods in the plant design, this paper provides validation analyses and results for the multiple-input analysis methods for piping by conducting excitation tests and validation analyses. Some recommendations were also found in the study for seismic design of NPPs.

Automatic machining system for the refurbishment of degraded welds in piping systems

Degradation of dissimilar metal welds in the piping systems of power plants is recognized as a significant ageing factor. Hence, close scientific attention is paid to various degradation mechanisms...

Thermocouple measurements to investigate the thermal fatigue of a cyclic thermal mixing process near a dissimilar weld seam

To investigate the thermal fatigue of piping systems in nuclear power plants, the Fluid-Structure-Interaction (FSI) test facility is operated at the University of Stuttgart. In the modular FSI-facility, a hot-water flow (265°C) in the main pipe (DN80) and a cold-water flow (20°C) in the branch pipe (DN40) are mixed within a horizontal mixing tee at 75bars. For the investigation of the impact of possible incomplete mixing on a dissimilar weld seam, a pipe section (module) with a dissimilar weld connection of an austenitic and a ferritic steel section is arranged near the mixing position. In order to achieve large temperature fluctuations the cold flow, which aims directly at the weld seam, is actively switched on and off by a valve-and-bypass system, while the hot flow is continuously maintained. The durations of the injection and bypass are varied with values of 30s (injection)/30s (bypass), 10s/10s, 5s/5s and 1s/1s.In this paper, the temperature fluctuations of these experiments are presented and analyzed in order to characterize the unsteady thermal loads on the weld materials. For this purpose, forty micro-thermocouples are installed in the weld seam module to measure the temperature in the near-wall fluid with a sampling rate of 100Hz. The distributions of the mean temperature and the temperature fluctuations in the flow within the weld seam module during the measurements are reconstructed. Fourier analyses are conducted and the results are summarized in the form of frequency spectra. Areas of the pipe wall with the highest potential for material damage due to thermal fatigue are identified with the help of these spectra. Furthermore, the possibility of thermal fatigue is confirmed by a metallographic investigation, in which a material crack is found.

A Feasibility Study of Noncontact Ultrasonic Sensor for Nuclear Power Plant Inspection

A pipe-wall thinning measurement is a key inspection to ensure the integrity of the piping system in nuclear power plants. To monitor the integrity of the piping system, a number of ultrasonic thickness measurements are manually performed during the outage of the nuclear power plant. Since most of the pipes are covered with an insulator, removing the insulator is necessary for the ultrasonic thickness measurement. Noncontact ultrasonic sensors enable ultrasonic thickness inspection without removing the insulator. This leads to reduction of the inspection time and reduced radiation exposure of the inspector. The inductively-coupled transducer system (ICTS) is a noncontact ultrasonic sensor system which uses electromagnetic induction between coils to drive an installed transducer. In this study, we investigated the applicability of an innovative ICTS developed at the University of Bristol to nuclear power plant inspection, particularly pipe-wall thinning inspection. The following experiments were performed using ICTS: thickness measurement performance, the effect of the coil separation, the effect of the insulator, the effect of different inspection materials, the radiation tolerance, and the measurement accuracy of wastage defects. These initial experimental results showed that the ICTS has the possibility to enable wall-thinning inspection in nuclear power plants without removing the insulator. Future work will address the issue of measuring wall-thinning in more complex pipework geometries and at elevated temperatures.

Transformation of fracture resistance curves by using bending modified Q (Qm) factor

The application of the leak-before-break (LBB) concept to design and assess piping systems for nuclear power plants requires fracture resistance curves (J-R curves) of cracked full scale pipes. However, it is expensive and difficult to perform full scale pipe tests. Laboratory test specimens with smaller geometry size were used to obtain the J-R curves for full scale pipes, but as both the specimen geometry and loading configuration have a strong effect on J-R curves, it is difficult to predict the J-R curves of full scale pipes accurately from laboratory test specimens. In this study, based on the bending modified Q (Qm) factor, and the J-R curves of compact tension (CT) specimens and curved wide plate (CWP) specimens, constraint corrected J-R curves of the stainless steel SA-312 TP304L were obtained, which were used to predict the J-R curves of full scale pipes. The predictions show good agreement with full scale pipe test results within the specific crack growth range.

NUMERICAL MODELING OF FLOW AND MASS TRANSFER EFFECT ON FLOW-ACCELERATED CORROSION IN CONSECUTIVE ELBOWS

Elbows are the most common fittings in piping systems of nuclear power plants, and can experience enhanced flow-accelerated corrosion (FAC) thinning rates in comparison to the straight pipe. When 2 elbows are in close proximity, the flow through the upstream elbow may influence the mass transfer and FAC thinning rate in the downstream elbow. Therefore, the flow condition and the corresponding mass transfer coefficient (MTC) determine the local distribution of FAC wall thinning in the elbows. This paper presents the numerical analysis of the MTC (in nondimensional form, Sherwood number) and its enhancement factor in small-bore piping (1″ inner diameter) under single-phase flow conditions, for 3 arrangements of 2 consecutive elbows: (i) in-plane S-configuration, (ii) in-plane C-configuration, and (iii) out-of-plane configuration. The commercial computational fluid dynamics tool ANSYS FLUENT® is used to simulate the flow and mass transfer for the various elbow configurations. The numerical results for the MTC enhancement factors for the 3 piping configurations agree with the experimental results. To apply these small-bore results to large-bore operating conditions found in nuclear power plants, the effect of Reynolds number is also considered. The results show that the MTC enhancement factors of the downstream elbow, affected by the upstream flow, is higher than that of the upstream elbow for all 3 consecutive elbow configurations.

Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of material degradation mechanisms and on effective strategies to manage the risks of debris-induced failures during a LOCA.

Thin-Plate-Type Embedded Ultrasonic Transducer Based on Magnetostriction for the Thickness Monitoring of the Secondary Piping System of a Nuclear Power Plant

Pipe wall thinning in the secondary piping system of a nuclear power plant is currently a major problem that typically affects the safety and reliability of the nuclear power plant directly. Regular in-service inspections are carried out to manage the piping system only during the overhaul. Online thickness monitoring is necessary to avoid abrupt breakage due to wall thinning. To this end, a transducer that can withstand a high-temperature environment and should be installed under the insulation layer. We propose a thin plate type of embedded ultrasonic transducer based on magnetostriction. The transducer was designed and fabricated to measure the thickness of a pipe under a high-temperature condition. A number of experimental results confirmed the validity of the present transducer.

CFX study of flow accelerated corrosion via mass transfer coefficient calculation in a double elbow

Using CFX numerical simulation, this study aims at calculating Mass transfer coefficient (MTC) as a parameter that indicate the FAC rate on the wall of the double elbow. A double elbow in the feeder water piping system of a typical combined cycle power plant that had been damaged due to the FAC was considered and was studied. In this study the influence of several parameters such as the mean flow velocity and the dimensionless distance between the two elbows (L/D) on the MTC at the elbows was investigated. The geometrical factor and the close proximity of two elbows with respect to the mass transfer coefficient were studied.It is found that mass transfer coefficient increases by Reynolds number. The maximum MTC was occurred in the intrados of second elbow. It is found that there is a very good coincidence between the locus in which the maximum MTC was calculated and the locus in which minimum wall thickness was measured.It is observed that as the remoteness between the elbows (L/D) increases, the Close Proximity Effect (CPE) is decreasing for L/D < 7. As L/D gets amplified (L/D > 7) the CPE is increasing very slowly and became constant. This CPE value is lower in comparison with that of L/D < 7. This is due to the fact that, when L/D > 7, the second bend experiences less impact due to the first bend and, also, there is a fully developed flow between both the bends.

Development of Wall-Thinning Evaluation Procedure for Nuclear Power Plant Piping—Part 1: Quantification of Thickness Measurement Deviation

Pipe wall thinning by flow-accelerated corrosion and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs). Most NPPs have management programs to ensure pipe integrity due to wall thinning that includes periodic measurements for pipe wall thicknesses using nondestructive evaluation techniques. Numerous measurements using ultrasonic tests (UTs; one of the nondestructive evaluation technologies) have been performed during scheduled outages in NPPs. Using the thickness measurement data, wall thinning rates of each component are determined conservatively according to several evaluation methods developed by the United States Electric Power Research Institute. However, little is known about the conservativeness or reliability of the evaluation methods because of a lack of understanding of the measurement error. In this study, quantitative models for UT thickness measurement deviations of nuclear pipes and fittings were developed as the first step for establishing an optimized thinning evaluation procedure considering measurement error. In order to understand the characteristics of UT thickness measurement errors of nuclear pipes and fittings, round robin test results, which were obtained by previous researchers under laboratory conditions, were analyzed. Then, based on a large dataset of actual plant data from four NPPs, a quantitative model for UT thickness measurement deviation is proposed for plant conditions.

Fracture studies on carbon steel elbows having part-through notch with and without internal pressure

Elbows are one type of components widely used in various industrial applications in conventional and nuclear power generating plants, chemical processing plants and petroleum refineries. Due to the complexities of geometric structure, elbows are one of the piping components that are prone to failure in the piping system of nuclear power plants (NPPs). In service, the piping components are subjected to internal pressure in addition to bending loads, and the internal pressure is known to have significant effect on the load carrying capacity of these components. Hence, fracture studies were carried out on 406 mm diameter elbows with and without internal pressure. The investigations have shown that the location of notch plays an important role in the load carrying capacity of the elbows and their failure behaviour. The internal pressure reduced the load carrying capacity of the elbow when the notch was located at the intrados in circumferential direction and the behaviour is vice-versa when the notch was located at the intrados in an axial direction, when subjected to opening moment, in both the cases.

多管体系での分岐部音響共鳴における湿り蒸気流中の圧力振幅特性

Flow-induced acoustic resonances in piping with closed side branches or T-junctions are one of the phenomena causing severe structural vibration and fatigue damage of piping and components in many engineering applications such as power plants. In the United States, the steam dryer of the boiling water reactor was damaged by high-cycle fatigue due to acoustic-induced vibration under a power uprating condition. Practical piping systems of power plants often have a steam flow, and moreover, the steam state can be not only dry steam but also wet steam, including nearly saturated conditions. Although many researchers have investigated acoustic resonances at side branches and T-junctions, it is considered that acoustic resonances under a wet steam flow have not yet been clarified since previous studies were mainly conducted under an air flow, and there have been few previous experiments performed under a steam flow, particularly a wet steam flow. The objective in this study is to investigate acoustic resonances in closed side branches with tandem- and coaxial-type arrangements under a wet steam flow. Experiments under a low-pressure condition for dry and wet steam flows were conducted and the results were compared. As a result, it is clarified that the effect of branch arrangements (single, tandem, coaxial) on the resonance amplitude under steam flow. In addition, differences in the amplitude response between dry and wet steam were evaluated.

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.

Non-contact monitoring of 3-dimensional vibrations of bodies using a neural network

Gas piping systems in power plants and factories are always influenced by the mechanical vibrations of rotational machines such as pumps, blowers, and compressors. Unusual vibrations in a gas piping system influence possible leakages of liquids or gases, which can lead to large explosive accidents. Real-time measurements of unusual vibrations in piping systems in situ prohibit them from being possible leakages owing to the repeated fatigue of vibrations. In this paper, a non-contact 3-dimensional measurement system that can detect the vibrations of a solid body and monitor its vibrational modes is introduced. To detect the displacements of a body, a stereoscopic camera system is used, through which the major vibration types of solid bodies (such as X-axis-major, Y-axis-major, and Z-axis-major vibrations) can be monitored. In order to judge the vibration types, an artificial neural network is used. The measurement system consists of a host computer, stereoscopic camera system (two-camera system, high-speed high-resolution camera), and a measurement target. Through practical application on a flat plate, the measured data from the non-contact measurement system showed good agreement with those from the original vibration mode produced by an accelerator.

Investigation on thermal stratification and turbulent penetration in a pressurizer surge line with an overall out-surge flow

Excessive thermal stresses caused by temperature fluctuations can result in thermal fatigue in the piping system of a nuclear power plants (NPPs). The phenomenon of thermal stratification and turbulent penetration in a pipe can lead to temperature fluctuations, and have attracted much attention of the researchers who deal with NPPs. In particular, this phenomenon in the pressurizer surge line has been assessed as one of the significant safety and technical issues due to its high rate of occurrence. This work focuses on a fundamental description of the thermal stratification and turbulent penetration in a pressurizer surge line using experiment and computational fluid dynamics (CFD) simulation based on the FLUENT 6.3.26 code. In this study, CFD simulation with the LES turbulence model is proved to be an accurate and effective method of studying thermal stratification in a pressurizer surge line. Details of turbulent penetration have also been studied in this paper. A criterion of determining the depth of turbulent penetration has been proposed, and the main frequency of the velocity fluctuations of turbulent penetration in a surge line is found to be a continuous frequency band, and not a fixed value.

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.

원주방향 표면 결함이 존재하는 배관에 가해지는 비틀림을 포함한 복합하중에 대한 한계하중식 제시

Since the Fukushima nuclear accident, several researchers are extensively studying the effect of torsion on the piping systems In nuclear power plants. Piping installations in power plants with a circumferential crack can be operated under combined loading conditions such as bending and torsion. ASME Code provides flaw evaluations for fully plastic fractures using limit load criteria for the structural integrity of the cracked pipes. According to the recent version of Code, combined loadings are provided only for the membrane and bending. Even though actual operating conditions have torsion loading, the methodology for evaluating torsion load is not established. This paper provides the results of limit load analyses by using finite element models for circumferentially cracked pipes under pure bending, pure torsion, and combined bending and torsion with tension. Theoretical limit load solutions based on net-section fully plastic criteria are suggested and verified with the results of finite element analyses.

THINNED PIPE MANAGEMENT PROGRAM OF KOREAN NUCLEAR POWER PLANTS

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion (FAC), cavitation, flashing and/or liquid drop impingements, are a main concern in carbon steel piping systems of nuclear power plant in terms of safety and operability. Thinned pipe management program (TPMP) had been developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning in the secondary side piping system. This program also consists of several technical elements such as prediction of wear rate for each component, prioritization of components for inspection, thickness measurement, calculation of actual wear and wear rate for each component. Decision making is associated with replacement or continuous service for thinned pipe components. Establishment of long-term strategy based on diagnosis of plant condition regarding overall wall thinning is also essential part of the program. Prediction models of wall thinning caused by FAC had been established for 24 operating nuclear plants. Long term strategies to manage the thinned pipe component were prepared and applied to each unit, which was reflecting plant specific design, operation, and inspection history, so that the structural integrity of piping system can be maintained. An alternative integrity assessment criterion and a computer program for thinned piping items were developed for the first time in the world, which was directly applicable to the secondary piping system of nuclear power plant. The thinned pipe management program is applied to all domestic nuclear power plants as a standard procedure form so that it contributes to preventing an accident caused by FAC.