• Journal of the Korean Society for Nondestructive Testing
• /
• v.34 no.5
• /
• pp.356-361
• /
• 2014

The eddy current testing performance directly affects the results of a steam generator tube integrity assessment because the integrity assessment of defected tubes is conducted based on eddy current testing results. This means that it may not be possible to accurately discriminate between adjacent flaws. This paper presents an investigation on the resolution of rotating pancake coils with multiple cracks and the effects on the structural integrity assessment of steam generator tubes.

• Nuclear Engineering and Technology
• /
• v.32 no.2
• /
• pp.99-107
• /
• 2000

A probabilistic integrity analysis method is presented for a reactor vessel under pressurized thermal shock(PTS) based on Monte Carlo simulation. This method can be applied to the structural integrity assessment of a reactor vessel subjected to pressurized thermal shock where the coolant temperature transient cannot be expressed explicitly as a time function. An axially or circumferentially oriented infinite length surface crack is assumed to be in the beltline weld region of the rector vessel's inside surface. The random variables are the initial crack depth, neutron fluence on the vessel's inside surface, the copper and nickel content of the vessel materials, R $T_{NDT}$ , $K_{IC}$ , and K/aub la/. The reliability of a sample reactor vessel under PTS is assessed quantitatively and the influence of the amount of neutron fluence is also examined by applying the present method.sent method.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.219.2-219
• /
• 1999

• Proceedings of the KSME Conference
• /
• 2007.05b
• /
• pp.3497-3500
• /
• 2007

The internal structure is subjected to dynamic analysis due to the structural integrity. The internal structure shall be installed in the vertical hole call IR1 of reactor core. In order to verify the deflection of the internal structure, the mode and response spectrum analysis of the internal structure was performed. The natural frequency of the internal structure is 11.6 Hz(mode 1 and 2) and deflections of the internal structure are less than values of allowable design (3.2 mm).

• Proceedings of the KSME Conference
• /
• 2008.11a
• /
• pp.264-268
• /
• 2008

An assessment of creep-fatigue crack behavior is required to ensure the structural integrity for high temperature components such as fast breeder reactor structures or thermal power plant components operating at an elevated temperature. In this study, an evaluation of creep-fatigue crack growth has been carried out according to the French assessment guide of the RCC-MR A16 for austenitic stainless steel structures. The assessment procedures for creep-fatigue crack growth in the recent version of the A16 (2007 edition) have been changed considerably from the previous version (2002 edition) and the material properties (RCC-MR Appendix A3) have been changed as well. The impacts of those changes on creep-fatigue crack growth behavior are quantified from the assessments with a structural model. Finally the assessment results were compared with the observed images obtained from the structural tests of the same structural specimen.

• Structural Monitoring and Maintenance
• /
• v.3 no.1
• /
• pp.51-69
• /
• 2016

Monitoring the performance and estimating the remaining useful life of aging civil infrastructure in the United States has been identified as a major objective in the civil engineering community. Structural health monitoring has emerged as a central tool to fulfill this objective. This paper presents a review of the major structural monitoring programs that have been recently implemented in the United States, focusing on the integrity and performance assessment of large-scale structural systems. Applications where response data from a monitoring program have been used to detect and correct structural deficiencies are highlighted. These applications include (but are not limited to): i) Post-earthquake damage assessment of buildings and bridges; ii) Monitoring of cables vibration in cable-stayed bridges; iii) Evaluation of the effectiveness of technologies for retrofit and seismic protection, such as base isolation systems; and iv) Structural damage assessment of bridges after impact loads resulting from ship collisions. These and many other applications show that a structural health monitoring program is a powerful tool for structural damage and condition assessment, that can be used as part of a comprehensive decision-making process about possible actions that can be undertaken in a large-scale civil infrastructure system after potentially damaging events.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.17 no.2
• /
• pp.131-140
• /
• 2004

It really needs fuzzy decision making of integrity assessment considering about both durability and load carrying capacity for maintenance and administration, such as repairing and reinforcing. This thesis shows efficient models about reinforced concrete structure using CART-ANFIS. It compares and analyzes decision trees parts of expert system, using the theory of fuzzy, and applying damage & diagnosis at reinforced concrete structure and decision trees of integrity assessment using established artificial neural. Decided the theory of reinforcement design for recovery of durability at damaged concrete ＆ the theory of reinforcement design for increasing load carrying capacity keep stability of damage and detection. It is more efficient maintenance and administration at reinforced concrete for using integrity assessment model of this study and can carry out predicting cost of life cycle.

• Proceedings of the KSME Conference
• /
• 2004.11a
• /
• pp.166-171
• /
• 2004

Since early 1950's 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, from which 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 working places. For this reason, a network system based on internet or intranet bas been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular 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 internet-based cooperative system for integrity evaluation system which adapts 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 agent programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet.

• Nuclear Engineering and Technology
• /
• v.55 no.4
• /
• pp.1365-1381
• /
• 2023

In this paper, a procedure for evaluating the structural integrity of the PCSG (Printed Circuit Steam Generator) unit block is presented with a simplified FE (finite element) analysis technique by applying the homogenization method. The homogenization method converts an inhomogeneous elastic body into a homogeneous elastic body with same mechanical behaviour. This method is effective when the inhomogeneous elastic body has repetitive microstructures, and thus the method was applied to the sheet assembly among the PCSG unit block components. From the method, the homogenized equivalent elastic constants of the sheet assembly were derived. The validity of the determined material properties was verified by comparing the mechanical behaviour with the reference model. Thermo-mechanical analysis was then performed to evaluate the structural integrity of the PCSG unit block, and it was found that the contact region between the steam header and the sheet assembly is a critical point where large bending stress occurs due to the temperature difference.

• Journal of Mechanical Science and Technology
• /
• v.20 no.12
• /
• pp.2061-2078
• /
• 2006

The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500$^{\circ}C$ and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated.