• 대한기계학회논문집A
• /
• 제29권7호
• /
• pp.947-954
• /
• 2005

The purpose of this study is to develop failure pressure evaluation models, which are applicable to straight pipes and elbows containing an internally wall thinning defect induced by flow-accelerated-corrosion (FAC). In this study, thus, three dimensional finite element (FE) analyses are performed to investigate the dependences of failure pressure of internally wall thinned pipe on the defect shape, the pipe geometry, and the defect location and bend radius of elbow. Also, the existing failure pressure assessment models for externally wall thinned pipes are examined. Based on these, the new models for assessing failure pressure of piping components with an internally wall thinning defect are proposed. Comparison of failure pressure, predicted by proposed models, with FE analysis result shows good agreement regardless of pipe type, defect shape, and defect location and bend radius.

• 한국압력기기공학회 논문집
• /
• 제11권2호
• /
• pp.51-60
• /
• 2015

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion, cavitation, flashing and/or liquid droplet impingement, is a main concern in secondary steam cycle piping system of nuclear power plants in terms of safety and operability. Thinned pipe management program (TPMP) has being developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning. In this paper, newest technologies, standards and regulations related to the integrity assessment, repair and replacement of thinned pipe component are reviewed. And technical improvement items in TPMP to secure the reliability and effectiveness are also presented.

• Earthquakes and Structures
• /
• 제17권5호
• /
• pp.489-499
• /
• 2019

Elastic-plastic behavior of nuclear power plant elbow piping under seismic loads has been conducted in this study. Finite element analyses are performed using classical Bilinear kinematic hardening model (BKIN) and Multilinear kinematic hardening model (MKIN) as well as a nonlinear kinematic hardening model (Chaboche model). The influence of internal pressure and seismic loading on ratcheting strain of elbow pipe is studied by means of the three models. The results found that the predicted results of Chaboche model is maximum, closely followed by the predicted results of MKIN model, and the minimum is the predicted results of BKIN model. Moreover, comparisons of analysis results for each plasticity model against predicted results for a equivalent cyclic loading elbow component and for a simplified piping system seismic test are presented in the paper.

• 한국압력기기공학회 논문집
• /
• 제11권2호
• /
• pp.13-19
• /
• 2015

Steam generator is one of the most important component of nuclear power plant, and it's integrity and reliability are to be assured to high level by pre-service inspection and in-service inspection. To improve the reliability of steam generator heat exchanger tubes and to secure the management of nuclear power plant safely, KHNP CRI recently has developed eddy current testing system for steam generator. KHNP CRI have performed a series of experimental verification and field application to confirm the performance of the developed ECT system in accordance with ASME Code requirements. The ECT system consists of a remote data acquisition unit, an ECT signal acquisition and analysis software, a water chamber robot controller and a probe push-puller. In this paper, we will details of the developed ECT system and the software and their experimental performance. And also we will report the field applying performance and the issues for further steps.

• Nuclear Engineering and Technology
• /
• 제52권12호
• /
• pp.2949-2957
• /
• 2020

Fatigue crack growth model has been developed for dissimilar metal weld joints of a piping component under cyclic loading, where in the crack is located at the center of the weld in the circumferential direction. The fracture parameter, Stress Intensity Factor (SIF) has been computed by using principle of superposition as K_H + K_M. K_H is evaluated by assuming that, the complete specimen is made of the material containing the notch location. In second stage, the stress field ahead of the crack tip, accounting for the strength mismatch, the applied load and geometry has been characterized to evaluate SIF (K_M). For each incremental crack depth, stress field ahead of the crack tip has been quantified by using J-integral (elastic), mismatch ratio, plastic interaction factor and stress parallel to the crack surface. The associated constants for evaluation of K_M have been computed by using the quantified stress field with respect to the distance from the crack tip. Net SIF (K_H + K_M) computed, has been used for the crack growth analysis and remaining life prediction by Paris crack growth model. To validate the model, SIF and remaining life has been predicted for a pipe made up of (i) SA312 Type 304LN austenitic stainless steel and SA508 Gr. 3 Cl. 1. Low alloy carbon steel (ii) welded SA312 Type 304LN austenitic stainless-steel pipe. From the studies, it is observed that the model could predict the remaining life of DMWJ piping components with a maximum difference of 15% compared to experimental observations.

• 한국압력기기공학회 논문집
• /
• 제14권2호
• /
• pp.49-58
• /
• 2018

The regulatory periodic inspection program (PSI) conducted at every overhaul period is the most important process for confirming the safety of nuclear power plants. The PSI for operating nuclear power plants in Korea mainly consist of component level performance check that had been developed based on deterministic approach putting the same degree of importance to all the inspection items. This inspection methodology is likely to be effective for preoperational inspection. However, once the plant is put into service, the PSI must be focused on whether to minimize the risk of accident using defense-in-depth concept and risk insight. The incorporation of defense-in-depth concept and risk insight into the deterministic based safety inspection has not been well studied so far. In this study, two track approaches are proposed to make sure that core damage be avoided: one is to secure success path and the other to block the failure path in a specific event tree of PSA. The investigation shows how to select safety important components and how to set up inspection group to ensure that core damage would not occur for a given initiating event, which results in strengthening defense-in-depth level 3.

• Nuclear Engineering and Technology
• /
• 제44권3호
• /
• pp.297-310
• /
• 2012

Aging-related degradation of nuclear power plant components is an important aspect to consider in securing the long term safety of the plant, especially the seismic safety, since the degradation of the components affects not only their seismic capacity but their response. This can cause a change in the seismic margin of a component and the overall seismic safety of a system. To better understand the status and characteristics of degradation of components in Nuclear Power Plants (NPPs), the degradation occurrences of components in the U.S. NPPs were identified by reviewing recent publicly available information sources and the characteristics of these occurrences were evaluated and compared to observations from the past. Ten categories of components that are of high risk significance in Korean NPPs were identified, comprising anchorage, concrete, containment, exchanger, filter, piping systems, reactor pressure vessels, structural steel, tanks, and vessels. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.

• 한국압력기기공학회 논문집
• /
• 제6권1호
• /
• pp.57-64
• /
• 2010

PHE(Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR(Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established the gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype to be tested in the loop. In this study, as part of the high-temperature structural-integrity evaluation of the PHE prototype, which is scheduled to be tested in the gas loop, we carried out high-temperature structural-analysis modeling, thermal analysis, and thermal expansion analysis of the PHE prototype. The results obtained in this study will be used to design the performance test setup for the PHE prototype.

• 한국압력기기공학회 논문집
• /
• 제5권1호
• /
• pp.25-31
• /
• 2009

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which comprises of the pressure boundary of primary system. The integrity of SG tube has been confirmed by the eddy current test every outage. In Korea, Bobbin probe and MRPC probe have been generally used for the eddy current test. Meanwhile the usage of Array probe has gradually increased in U.S., Japan and other countries. In this study, we investigated the defect detection capability of the Array probe through its preliminary application to SG tube inspection. The Array probe has the equivalent capability in the defect detection and sizing as the conventional methods. Thus it is desirable that the Array probe is generally applied to SG tube inspection in the domestic NPPs.

• 한국압력기기공학회 논문집
• /
• 제8권1호
• /
• pp.1-7
• /
• 2012

A PHE (Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X is being tested in a small-scale gas loop at Korea Atomic Energy Research Institute. In order to properly evaluate the high-temperature structural integrity of the small-scale PHE prototype, it is very important to impose a proper constraint condition on its structural analysis model. For this effort, we tried to impose several constraint conditions on the structural analysis model and consequently fixed a proper and effective displacement constraints.