Youn Won Park

Determination of equivalent single crack based on coalescence criterion of collinear axial cracks

In a nuclear power plant the steam generator tubes cover a major portion of the primary pressure-retaining boundary. Thus, very conservative approaches have been taken in the light of steam generator tube integrity. According to the present criteria, tubes wall-thinned in excess of 40% should be plugged whatever the cause. However, many analytical and experimental results have shown that no safety problems exist even with thickness reductions greater than 40%. The present criterion was developed about 20 years ago when wear and pitting were dominant causes for steam generator tube degradation, and it is based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is more common in general. The objective of this study is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence model for two adjacent through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we reviewed the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence model, we performed finite element analysis and some parametric studies. Then, we developed a coalescence evaluation diagram.

Investigation on the interaction effect of two parallel axial through-wall cracks existing in steam generator tube

It is commonly required that steam generator tubes wall-thinned in excess of 40% should be plugged. However, the plugging criterion is known to be too conservative for some locations and types of defects and its application is confined to a single crack. In the previous study, the conservatism of the present plugging criterion of steam generator tubes was reviewed and a crack coalescence model applicable to steam generator tubes with two collinear axial through-wall cracks was proposed. Since parallel axial cracks are more frequently detected during in-service inspections than collinear axial cracks, the studies on parallel axial cracks spaced in circumferential direction are necessary. The objective of this paper is to investigate the interaction effect between two parallel axial through-wall cracks existing in a steam generator tube. Finite element analyses were performed and a new failure model of the steam generator tube with these types of cracks is suggested. Interaction effects between two adjacent cracks were investigated to explain the deformation behavior of cracked tubes.

A study on technique to estimate impact location of loose part using wigner-ville distribution

Presented in this paper is a method to estimate impact location of a loose part using the Wigner-Ville distribution. The method uses dispersion characteristics of bending waves propagated in a plate. The power propagation velocity and arrival time difference of bending waves related to the dispersion characteristics can be obtained through the transformation of impact signals using the Wigner-Ville distribution. The distance from the impact location to the signal measuring point can be estimated using the information on the power propagation velocity and the arrival time difference of two bending waves. The experimental results show that the proposed method estimates the impact location with relative percentage error within 10% compared with the actual impact location.

Pressurized thermal shock analyses of a reactor pressure vessel using critical crack depth diagrams

Evaluated in this study is the pressure vessel integrity under a pressurized thermal shock. Using transient histories such as temperature, pressure and heat transfer coefficient, the stress distribution is calculated and then stress intensity factors are obtained for a wide range of crack sizes. The stress intensity factors are compared with the fracture toughness to check if cracking is expected to occur during the transient. Critical crack depth diagrams are prepared for each transient which is expected to initiate a pressurized thermal shock accident. Plant-specific analyses of the most limiting plant in Korea are performed to assure the structural integrity of the reactor vessel and the results are discussed.

Round robin analysis of pressurized thermal shock for reactor pressure vessel

A comparative assessment study is performed here for deterministic fracture mechanics analysis of a pressurized thermal shock (PTS). A round robin problem is proposed using the data available in Korea and all organizations interested in the PTS analysis are invited. The problems consisting of two transients and 10 cracks are solved and their results are compared to generate a reference solution that could serve as benchmarks for future qualification of analytical method. Nine participants from seven organizations responded to the problem and their results are compiled in this paper.

A PROBABILISTIC INTEGRITY ASSESSMENT OF FLAW IN ZIRCONIUM ALLOY PRESSURE TUBE CONSIDERING DELAYED HYDRIDE CRACKING

In the CANDU nuclear reactor, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the nuclear fuel bundles and heavy water coolant. Pressure tubes are major component of nuclear reactor, but only selected samples are periodically examined due to numerous numbers of tubes. Pressure tube material gradually pick up deuterium, as such are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC), which is the characteristic of pressure tube integrity evaluation. If cracks are not detected, such a cracking mechanism could lead to unstable rupture of the pressure tube. Up to this time, integrity evaluations are performed using conventional deterministic approaches. So it is expected that the results obtained are too conservative to perform a rational evaluation of lifetime. In this respect, a probabilistic safety assessment method is more appropriate for the assessment of overall pressure tube safety. This paper describes failure criteria...

Development of Web-Based Fatigue Life Evaluation System for Reactor Pressure Vessel

While the demand on electric power is consistently increasing, public concerns and regulations for the construction of new nuclear power plants are getting restrict, and also operating nuclear power plants are gradually ageing. For this reason, the interest on lifetime extension for operating nuclear power plants by applying lifetime management system is increasing. The 40-year design life concept was originally introduced on the basis of economic and safety considerations. In other words, it was not determined by technological evaluations. Also, the transient design data which were applied for fatigue damage evaluation were overly conservative in comparison with actual transient data. Therefore, the accumulation of fatigue damage may result in a big difference between the actual data and the design data. The lifetime of nuclear power plants is mostly dependent on the fatigue life of a reactor pressure vessel, and thus, the exact evaluation of fatigue life on a reactor pressure vessel is a crucial factor in determining the extension of operating life. The purpose of this paper is to introduce a real-time fatigue monitoring system for an operating reactor pressure vessel which can be used for the lifetime extension. In order to satisfy the objectives, a web-based transient acquisition system was developed, thereby, real-time thermal-hydraulic data were reserved for 18 operating reactor pressure vessels. A series of finite element analyses was carried out to obtain the stress data due to actual transient. The fatigue life evaluation has been performed based on the stress analysis results and, finally, a web-based fatigue life evaluation system was introduced by combining analysis results and on-line monitoring system. Comparison of the stress analysis results between operating transients and design transients showed a considerable amount of benefits in terms of fatigue life. Therefore, it is anticipated that the developed web-based system can be utilized as an efficient tool for fatigue life estimation of reactor pressure vessel.

Statistical Reliability Assessment of UT Round-Robin Test Data for Piping Welds

Ultrasonic NDE is one of important technologies in the life-time maintenance of nuclear power plant. Ultrasonic inspection system is consisted of the operator, equipment and procedure. The reliability of ultrasonic inspection system is affected by its ability. The performance demonstration round robin was conducted to quantify the capability of ultrasonic inspection for in-service. Several teams employed procedures that met or exceeded with ASME Sec. XI code requirements detected the piping of nuclear power plant with various cracks to evaluate the capability of detection and sizing. In this paper, the statistical reliability assessment of ultrasonic nondestructive inspection data using probability of detection (POD) is presented. The result of POD using logistic model was useful to the reliability assessment for the NDE hit or miss data.

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Reactor Pressure Vessel

Since early 1950’s, the fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, and as a result, various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of locations. For this reason, a network system based on internet or intranet bas been appeared in various fields of business. Evaluating the integrity of critical components is one of the most critical issues in the nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including periodical in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adopts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses virtual reality (VR) technique, virtual network computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and provide experts to co-operate each other by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel. Introduction Since the British Calder Hall nuclear power plant has started its operation in 1956 for the first time in the world, nuclear power plants have been major resources for power generation for last 50 years. The safety of a nuclear power plant, however, has seriously been considered as a threatening factor to the environment due to the detrimental damage in case of accidents. In order to prevent fatal accidents, nuclear power plants are built and managed under strict regulations. There have been a number of accidents reported so far. One of the major issues is a defect assessment which is contained in major components. Major components of a nuclear power plant are regularly inspected with non-destructive testing equipments. If a defect is found during the inspection, it should be evaluated by strict regulatory codes, such as ASME Sec. XI[1]. These regulatory codes usually include complicated calculation procedures All correspondences are to be sent to Professor Choi at boong33@skku.edu, Fax: +82-31-290-5276 Key Engineering Materials Online: 2004-08-15 ISSN: 1662-9795, Vols. 270-273, pp 2244-2249 doi:10.4028/www.scientific.net/KEM.270-273.2244