Nuclear Piping Research Articles

A probabilistic approach for estimating defect size and density considering detection uncertainties and measurement errors

Every structural component has some form and amount of hidden defects, however small it might be. These defects should be detected and characterized in order to ensure reliability of safety-critical structures. Test equipment is often used to inspect and gather data from probable defect sites in structural components during in-service inspections for maintenance purposes. This data is then analyzed and processed to measure the defect size and density. In-service inspection data may be highly uncertain in nature owing to detection uncertainties and measurement errors typically associated with the testing and evaluation process. These uncertainties and errors, if not properly accounted for, can result in defect size and density estimates not representative of the true state of degradations. In this article we first provide a description of the detection uncertainties and measurement errors in the context of nondestructive testing and evaluation. We then propose a Bayesian approach that updates prior knowledge of defect size and density with uncertain nondestructive evaluation data, accounting for detection uncertainties, measurement errors, and other associated uncertainties, to infer the posterior estimates of true size and number of defects. An example application of the proposed approach is then presented for estimating true defect size and density in nuclear reactor piping welds using ultrasonic evaluation data. The Bayesian approach presented in this article fills a critical gap in health management and reliability prognosis of structures, thereby enhancing the overall structural safety and operability.

Low Cycle Fatigue Behavior and Seismic Assessment for Elbow Pipe Having Local Wall Thinning

One of the concerned technical issues in the nuclear piping under operation is pipe wall thinning caused by flow accelerated corrosion. Recently, it has been reported that the elbow section is more suspicious on pipe wall thinning by erosion–corrosion. Some researchers including authors have been studied static and fatigue strength of elbows with local wall thinning. However, still more experiment and analysis data are needed to clarify the technical issues. Accordingly, further experiments and their evaluations were carried out by the authors. This paper presents the influences of size and location on fatigue life. Also as one of the application of the test results, safety margin of elbows with wall thinning against seismic loading is discussed. Low cycle fatigue tests were conducted using elbow specimens made of STPT410 steel with local wall thinning. The local wall thinning was machined on the inside of elbow specimens in order to simulate erosion/corrosion metal loss. The local wall thinning areas were located at three different areas, called extrados, crown, and intrados. Eroded ratio (eroded depth/wall thickness) is 0.5 and 0.8 and eroded angle is 90 deg and 180 deg. The elbow specimens were subjected to cyclic in-plane bending under displacement control (±20 mm) without and with internal pressure of 3 MPa using a universal testing machine. Fatigue life was defined as fatigue crack penetration through the thickness and crack penetration was watched by naked eyes during the test through the protection window made of a transparent plastic plate. Obtained main conclusions are as follows: (1) Existence of local wall thinning in extrados does not have an important effect on fatigue life. Especially, fatigue crack does not initiate at the extrados where the extreme local wall thinning exists (eroded ratio = 0.8 and eroded angle = 180 deg). (2) Regardless of existence of internal pressure, fatigue crack initiates at the crown where local wall thinning does not exist for an elbow with local wall thinning at extrados. This conclusion should be confirmed using eroded elbow specimens under more high pressure. (3) Even if the eroded ratio and the eroded angle reached up to 0.8 and 180 deg, the elbows with local wall thinning have high safety margin against seismic loading, comparing to ASME Boiler and Pressure Vessel Code Section 3 allowable seismic stress criteria.

Effects of radial gap and penetration depth on vibration fatigue behaviour of 304L stainless steel piping with socket weld

Fatigue related failures of small bore piping caused by vibration of socket welds often occur at nuclear power plants. The purpose of the present study is to evaluate the integrity of the socket weld in nuclear piping under the condition of vibration testing. The specimens were bolted to the shaker table and shaken simultaneously near their resonant frequencies to produce the desired stress. The test results show that the failures at higher stress tend to originate at the toe, while that for the case of lower stress failures tends to occur at the root. The effects of penetration depth and radial gap were also investigated by finite element method. Results show that the higher penetration depth and radial gap can decline the stress distribution in the weld root, which is beneficial to fatigue life. Further, the desired radial gaps can significantly reduce the von Mises stresses, which decrease from 150·4 to 82·7 MPa with the width of radial gap from 0·1 to 0·4 mm.

Infrared Thermography Characterization of Defects in Seamless Pipes Using an Infrared Reflector

Infrared thermography uses infrared energy radiated from any objects above absolute zero temperature, and the range of its application has been constantly broadened. As one of the active test techniques detecting radiant energy generated when energy is applied to an object, ultrasound infrared thermography is a method of detecting defects through hot spots occurring at a defect area when 15~100 kHz of ultrasound is excited to an object. This technique is effective in detecting a wide range affected by ultrasound and vibration in real time. Especially, it is really effective when a defect area is minute. Therefore, this study conducted thermography through lock-in signal processing when an actual defect exists inside the austenite STS304 seamless pipe, which simulates thermal fatigue cracks in a nuclear power plant pipe. With ultrasound excited, this study could detect defects on the rear of a pipe by using an aluminium reflector. Besides, by regulating the angle of the aluminium reflector, this study could detect both front and rear defects as a single infrared thermography image.

B-Scan 초음파 측정장비를 이용한 원전 배관 침식손상 검사법 개발

B-Scan 초음파 측정장비를 이용한 원전 배관 침식손상 검사법 개발

On the Shape of Stress Corrosion Cracks in Water-Cooled Nuclear Power Reactor Piping

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Statistical characterisation of B2 for nuclear pipe bends: In-plane closing moment

In recent years, the necessity of reliability-based design has received significant attention from researchers working in the area of pressure vessels and nuclear piping. Preliminary guidelines have been formulated for the reliability-based design of straight pipes using the load and resistance factor design (LRFD) format. In order to formulate similar LRFD design equations for pipe bends and elbows, the first step is a statistical characterisation of the stress indices B1 and B2, which play a significant role for the stress analysis of bend geometry. This paper focuses on the statistical characterisation of B2, which arises due to fabrication/geometric uncertainties, for a thin stainless steel pipe with a 90° bend, subjected to in-plane closing moment. Based on Monte Carlo simulation and nonlinear finite element analyses, the code-based expression for B2 is revised with a random variable K replacing the constant numerator. The statistics for K are provided for different pipe nominal dimensions. The overall mean value of K indicates that the existing code provision gives a conservative estimate of plastic collapse moment for pipe bends subjected to in-plane closing bending moments.

전단간섭계를 이용한 직관시험편의 결함 깊이 측정

Abstract In the nuclear industry, wall thinning defect of straight pipe occur the enormous loss in life evaluation and safety evaluation. To use non-destructive technique, we measure deformation, vibration, defect evaluation. But, this techniques are a weak that is the measurement of the wide area is difficult and the time is caught long. In the secondary side of nuclear power plants mostly used steel pipe, artificiality wall thinning defect make in the side and different thickness make to the each other. wall thinning defect part of deformation measure by using shearography. In addition, optical measurement through deformation, vibration, defect evaluation evaluate pipe and thickness defects of pressure vessel is to evaluate quantitatively. By shearography interferometry to measure the pipe's internal wall thinning defect and the variation of pressure use the proposed technique, the quantitative defect is to evaluate the thickness of the surplus. The amount of deformation use thickness of surplus prediction of the actual thickness defect and approximately 7 percent error by ensure reliability. According to pressure the amount of deformation and the thickness of the surplus through DB construction, nuclear power plant pipe use wall thinning part soundness evaluation. In this study, pressure vessel of thickness defect measure proposed nuclear pipe of wall thinning defect prediction and integrity assessment technology development. As a basic research defected theory and experiment, pressure vessel of advanced stability and soundness and maintainability is expected to contribute foundation establishment.

Design, calibration and testing of a thin film temperature gauge array for temperature and heat flux measurements in fluid mixing experiments

Material aging (“thermal fatigue”) in nuclear power plant piping systems due to thermal fluctuations has been identified as a high ranking safety issue (NEA/CSNI/R(2007)13 (2008) [19]). In order to provide a reliable assessment of potential material stresses the temperature fluctuations need to be well characterized as an input entity for analytical investigations. This contribution presents a sensor able to characterize thermal fluctuations on the wall—an entity necessary as input to analytical investigation.An array of resistive temperature detectors (RTDs) based on a micro patterned thin film platinum resistor has been built. The sensors are located in a glass substrate (40×53 mm2), in which the platinum resistors and the electrical contact leads are embedded. The operational temperature range is below 0 °C up to 200 °C, the temperature coefficient of resistivity (TCR) is 0.0011 °C−1 with a high correlation coefficient, and the thermal time constant has been verified at about 30 ms. Such sensors can be used to measure and characterize in terms of frequency and amplitude of thermal fluctuations at the inner wall of T-junction pipelines in order to evaluate the heat flux through the wall and thus the heat transfer coefficient.

원전 배관 감육 결함 검사를 위한 IR 열화상시험 조건 결정

This study conducted infrared (IR) thermography tests using pipe and plate specimens with artificial wall-thinning defects to find an optimal condition for IR thermography test on the wall-thinned nuclear piping components. In the experiment halogen lamp was used to heat the specimens. The distance between the specimen and the lamp and the intensity of halogen lamp were regarded as experimental parameter. When the distance was set to 1~2 m and the lamp intensity was above 60 % of full power, a single scanning of IR thermography detected all artificial wall-thinning defects, whose minimum dimension was $2{\Theta}

OS0806 原子力配管の耐震終局強度評価法の開発(1) : 地震荷重に対する限界状態関数

In order to develop realistic failure evaluation method for nuclear power plant piping, the past results of study were investigated. Based on them, the dominant failure mode against seismic load was revealed, and the limit state function to the mode was proposed.

The Compaoparison of Characteristics of Fracture Toughness of Materials for Nuclear Piping by Use of Various Types of Specimens

Since it is difficult to cut off large specimens from nuclear piping, generally the standard CT specimens of 1T are used to perform fracture toughness tests. However, when compared with the test result obtained from a real pipe, the results from CT specimens are been known to be very conservative. Therefore, the U.S. Nuclear Regulatory Commission recommends the use of specimens whose thickness is greater than or equal to the thickness of a real pipe for estimating the stability of crack resistance[1]. In this study, a series of fracture toughness tests were performed using various types of specimens: the standard CT, non-standard Compact Pipe, and Curved CT. The results of the tests were compared.

Evaluation of fracture toughness characteristics for nuclear piping using various types of specimens

The leak-before-break (LBB) concept is based on the fracture resistance curve obtained by J-R tests for various types of specimens. But fracture toughness data of various types of specimens are different. For example, the estimates for fracture toughness of standard specimens are conservative when compared to the fracture toughness of real pipes because of the difference of the constraint effect between real pipes and standard specimens. Therefore, to estimate the integrity of real pipes based on elastic–plastic fracture mechanics, we have to first establish the relations among real pipes, standard CT specimens and curved CT specimens. In this study, we perform fracture toughness tests of nuclear piping materials using various types of specimens. And each J-R curve considered in the constraint effect is compared. The experimental results showed that a curved CT specimen is similar to the cases of pipes with small crack angles (60° and 120°) and a CT specimen is similar to the case of pipes with the crack length of 180°. Also, the Q-stress for curved CT specimens is higher than that of pure bending pipe specimens and lower than that of CT specimens.

Study on analogy between dynamic load and displacement-controlled load

Difficulties in handling dynamic load with elastic–plastic analysis and/or structural component test significantly exist in the field of R&D for seismic resistive nuclear pipe systems. In previous studies, the analyses and the structural tests were conducted statically. Many of them have been suggesting conservativeness included in the current seismic criterion of the nuclear design code. Therefore, if analogy between the earthquake load and static displacement are proved, it should bring possibility of more reasonable criterion. The author studied analogy between dynamic load and static displacement using FEM analysis with elastic beam model and elastic–plastic shell model of an elbow. First, the dynamic load equivalent to 160% of ASME/JSME Code’s Level D limit was defined using the elastic beam model. Next, the defined dynamic load was given to the elastic–plastic shell model to acquire the non-linear behaviors such as the relative displacement trend and the plastic strain accumulation at crown of the elbow. Then, three cases of cyclic displacement trends, Case 1, 2 and 3, were defined base on the dynamic analysis results and given to the elastic–plastic shell model statically. Comparison between the dynamic analysis and the static analysis of Case 1 showed clear analogy between the dynamic load and the static displacement. Case 2, however, was found to be difficult to be judged at this moment. Case 3 was conservative throughout the loaded time and suggested possibility of the conservative displacement trend for static structural test without conducting the elastic–plastic analysis.

Application of piping failure database to nuclear safety issues in Korea

This paper describes analyses and application of OPDE (OECD Piping Data Failure Exchange) database and NPFD (Nuclear Piping Failure Database) to some safety related issues for Korea nuclear power plants. Safety issues considered in this study include the piping rupture frequency assessment, the screening criteria of the leak before break application, and the augmented in-service inspection and the risk-informed ISI (RI-ISI). As a result of this study: (1) Piping rupture frequencies of the very small LOCA, feedwater line break events, and flood events were evaluated and used as initiating event frequencies for the PSA of Korean PWR plants; (2) Based on the review of OPDE DB and LBB screening criteria, LBB application to the pressurizer surge line, safety injection system, and shutdown cooling system were approved under the conditions that detailed thermal stress analyses should be performed to the systems. (3) For RI-ISI application, major degradation mechanisms in the reactor coolant system were analyzed based on the analysis of the OPDE and NPFD and augmented inspections were enforced to the pipe segments susceptible to degradation mechanism such as thermal stratification/fatigue.

Studies on the effect of compliance on fracture behaviour of carbon steel pipes with circumferential through-wall crack

The design of Nuclear Power Plant piping is based on Leak-Before-Break (LBB) design philosophy. The integrity of cracked nuclear piping system in LBB is often examined from conventional fracture mechanics. Conventionally the fracture analysis is performed by assuming infinite compliance of the piping system, which means that the stiffening effect of the connected uncracked piping system is fully neglected. However, in reality the ends of the pipe are connected to the rigid pressure vessels, primary coolant pipes, etc. In order to study the effect of compliance on the fracture behaviour of pipes, fracture studies were carried out on fourteen specimens of 219 mm outer diameter carbon steel pipes having circumferential through-wall crack. The fracture tests were conducted under four point bending. Out of the fourteen specimens, twelve specimens were tested in fixed end condition (finite compliance) and the remaining two specimens were tested in simply supported condition (infinite compliance). These pipes were made of SA 333 Gr. 6 carbon steel. Extensive instrumentation was carried out on these pipes to collect various data, e.g., load, load-line displacement (LLD), crack mouth opening displacement (CMOD), rotation at the ends (flanges), deflection at critical locations and surface crack growth. Surface crack growth was monitored using Image Processing Technique (IPT). The investigations have shown that the load carrying capacity of the pipes under finite compliance condition (fixed end condition) increases when compared to infinite compliance (simply supported condition). The load carrying capacity of the pipes under finite compliance weakly depends on the initial crack angle where as the load carrying capacity of the pipes under infinite compliance strongly depends on the initial crack angle. The safety margin is quite large particularly with a stiff piping system. This benefit in terms of increase in safety margin from LBB point of view is due to the moment redistribution. This paper discusses the effects of notch angle, span and the support conditions on the fracture behaviour of pipes with circumferential through-wall crack.

Cyclic fracture behaviour of 304LN stainless steel under load and displacement control modes

ABSTRACTAttempts have been made to understand cyclic fracture behaviour of AISI 304LN stainless steel used for nuclear piping materials under load vis‐à‐vis displacement controlled fracture tests; the former closely simulate the seismic loading conditions. The load controlled tests indicate that a material fails in a limited number of cycles even when the load amplitudes are sufficiently below the maximum load in a monotonic J–R test. The displacement controlled tests, on the other hand, show that the energy absorbing ability of a material gets severely reduced under cyclic loading conditions. The obtained results on standard laboratory specimens have been compared with similar available results on components in order to provide guidelines for maximum load bearing capability of engineering components under cyclic loading.

Estimates of Mechanical Properties and Residual Stress of Narrow Gap Weld for Leak-Before-Break Application to Nuclear Piping

The present paper addresses the mechanical characteristics of the stainless steel narrow gap weld (NGW) for the leak-before-break (LBB) application to the main loop piping of a nuclear power plant. Recently, in Korea, the connection with the reactor coolant main loop piping and the steam generator has been welded with ER308L NGW after the replacement of a steam generator of a Korean nuclear power plant. The NGW technique has many merits, for instance, the reduction of construction time and the reduction of shrinkage and deformation after welding due to its small groove angle and welding bead width compared with the conventional welds. In this paper, the tensile and fracture toughness test results of the three ER308L test coupons from NGW were presented and compared with those from conventional welds at the operating condition of the nuclear power plant. In addition, the distribution of the welding residual stress as well as the deformation behavior of the ER308L weld due to NGW was predicted through the nonlinear two-dimensional finite element analysis in which the detailed actual welding process of NGW was simulated. The results presented in this paper can be used to evaluate LBB application to nuclear piping with NGW and to provide the important information to perform the flaw evaluation as well as improve the weld procedure of NGW.

Recommended Revisions to Nuclear Piping Thermal Expansion Stress Limits

Piping thermal expansion stress limits for ASME Section III Class 2 and 3 piping are directly based on the original thermal expansion rules for B31.1 piping published in 1955. Temperature limits for B31 piping are much higher than for Class 2/3 nuclear piping. The fundamental concern in 1955 for piping design was to prevent cyclic creep strains, and the thermal expansion rules were based on a strain range criterion. Fatigue was also a concern, and a stress-range reduction factor for cyclic conditions was incorporated into the strain-based rule. The history and development of the B31.1 thermal expansion rules are summarized. After review of the technical basis, it becomes evident that the nuclear Class 2/3 rules can be simplified and improved by changing to an explicit fatigue-based rule rather than the strain-based rule of B31.1. Recommended Section III code revisions are provided with discussion of related technical issues.

Reliability analysis of nuclear piping system using semi-Markov process model

The paper presents a general model for evaluating the rupture frequencies and reliability of the piping system in nuclear power plant based on the theory of semi-Markov process. The proposed model is able to incorporate the effect of aging related degradation of pipes. Time dependent rupture frequencies are computed and compared against those obtained from the homogeneous Markov process model. The influence of flaw and leak inspection intervals on the piping rupture probability is investigated.