ABSTRACT: The 40% of wall thickness criterion which has been used as a plugging rule is applicable only to a single cracked steam generator
tubes. In the previous studies performed by authors, several failure prediction models were introduced to estimate the plastic
collapse pressures of steam generator tubes containing two adjacent collinear or parallel axial through-wall cracks. The objective
of this study is to examine the failure prediction models and propose optimum ones for two non-aligned axial through-wall
cracks in steam generator tubes. In order to determine the optimum ones, a series of plastic collapse tests and finite element
analyses were carried out for steam generator tubes with two machined non-aligned axial through-wall cracks. Thereby, either
the plastic zone contact model or COD based model was selected as the optimum one according to axial distance between two
cracks. Finally, the optimum failure prediction model was used to demonstrate the conservatism of flaw characterization rules
for multiple cracks having equal lengths according to ASME code.
International Journal of Fracture 04/2012; 144(2):91-101. · 1.49 Impact Factor
ABSTRACT: The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle. In order to complete the
integrity evaluation of pressure tube, expert knowledge, iterative calculation procedures and a lot of input data are required.
More over, results of integrity assessment may be different according to the evaluation method. For this reason, an integrity
evaluation system, which provides efficient way of evaluation with the help of attached database, was developed. The present
system was built on the basis of 3D FEM results, ASME Sec. XI, and Fitness For Service Guidelines for CANDU pressure tubes
issued by the AECL (Atomic Energy Canada Limited). The present system also covers the delayed hydride cracking and the blister
evaluation, which are the characteristics of pressure tube integrity evaluation. In order to verify the present system, several
case studies have been performed and the results were compared with those from AECL. A good agreement was observed between
those two results.
Journal of Mechanical Science and Technology 01/2003; 17(7):947-957. · 0.45 Impact Factor
ABSTRACT: 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 for probabilistic analysis and fracture mechanics analyses of the pressure tubes in consideration of DHC. Major input parameters such as initial hydrogen concentration, the depth and aspect ratio of an initial surface crack, DHC velocity and fracture toughness are considered as probabilistic variables. Failure assessment diagram of pressure tube material is proposed and applied in the probabilistic analysis. In all the analyses, failure probabilities are calculated using the Monte Carlo simulation. As a result of analysis, conservatism of deterministic failure criteria is showed.
International Journal of Modern Physics B - IJMPB. 01/2003; 17:1587-1593.
ABSTRACT: Structural and mechanical components in a nuclear power plant are designed to operate for its entire service life. Recently, a number of nuclear power plants are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. However, only a small portion of these components are of great importance for a significant aging degradation which would deeply affect the long-term safety and reliability of the related facilities. Therefore, it is beneficial to build a monitoring system to measure an aging status. While a number of integrity evaluation systems have been developed for NPPs, a real-time aging monitoring system has not been proposed yet
 and . This paper proposes an expert system for the integrity evaluation of nuclear power plants based on a Web-based Reality Environment (WRE). The proposed system provides the integrity assessment for the major mechanical components of a nuclear power plant under concurrent working environments. In the WRE, it is possible for users to understand a mechanical system such as its size, geometry, coupling condition etc. In conclusion, it is anticipated that the proposed system can be used for a more efficient integrity evaluation of the major components subjected to an aging degradation.
International Journal of Pressure Vessels and Piping 87(1):33-40. · 0.99 Impact Factor
ABSTRACT: 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.
Nuclear Engineering and Design. 214:13-23.
ABSTRACT: 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.
Nuclear Engineering and Design.