• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.61-67
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• 2015

3D FEM modeling based on 3D CAD data has been performed to evaluate the efficiency of CP system in a real operating nuclear power plant. The results of it successfully produced sophisticated profiles of electrolytic potential and current distributions in the soil of an interested area. This technology is expected to be a breakthrough for detection technology of damages on buried pipes when it comes into combining with a brand of area potential earth current (APEC) and ground penetrated radar (GPR) technologies. 2D current distribution and 2D current vectors on the earth surface from the APEC survey will be used as boundary conditions with exact 3D geometry data resulting in visualization of locations and extents of corrosion damages on the buried pipes in nuclear power plants.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.35-46
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• 2020

In this paper, a fracture mechanics evaluation system which can be used to assess the leak-before-break (LBB) of nuclear piping is developed. Existing solutions for calculating the fracture mechanics parameters (J-integral and crack opening displacement) required for LBB evaluation were firstly presented. Then a module for calculating J-integral and COD was developed, with an additional module for predicting the critical load based on the crack driving force diagram to finally develop a fracture mechanics evaluation system. To confirm the validity of the proposed evaluation system, finite element (FE) analysis was performed, and the FE J-integral and COD results were compared with prediction results using the J-integral and COD estimations program. Furthermore, the critical load assessment module was verified by comparing the actual pipe test results (Battelle test data) with prediction results using the proposed program.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.380-385
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• 2010

The integrity of components and piping of operating nuclear power plants has been identified by in-service inspection(ISI) requirements and activities commensurate with standards and codes such as KEPIC MI or ASME Code Section XI. However, the other various degradation mechanisms not considered during design stage of nuclear power plants have been checked by enhanced ISI. The requirements of enhanced ISI have been voluntarily developed by the industry itself or strickly issued by regulatory body. Even though the requirements were developed by the industry, they should be reviewed by regulatory body for their application in nuclear power plants. The enhanced ISI activities and requirements of non-destructive examination(NDE) which reflect the degradation issues in nuclear power industry will be primarily discussed in this paper.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.32 no.1
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• pp.17-27
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• 2019

In this paper, we study the applicability of Tuned Mass Damper(TMD) to improve seismic performance of piping system under earthquake loading. For this purpose, a mode analysis of the target pipeline is performed, and TMD installation locations are selected as important modes with relatively large mass participation ratio in each direction. In order to design the TMD at selected positions, each corresponding mode is replaced with a SDOF damped model, and accordingly the corresponding pipeline is converted into a 2-DOF system by considering the TMD as a SDOF damped model. Then, optimal design values of the TMD, which can minimize the dynamic amplification factor of the transformed 2-DOF system, are derived through GA optimization method. The proposed TMD design values are applied to the pipeline numerical model to analyze seismic performance with and without TMD installation. As a result of numerical analyses, it is confirmed that the directional acceleration responses, the maximum normal stresses and directional reaction forces of the pipeline system are reduced, quite a lot. The results of this study are expected to be used as basic information with respect to the improvement of the seismic performance of the piping system in the future.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.389-396
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• 2018

The Kori-1 Nuclear Power Plant (NPP), WH 2-Loop Pressurized Water Reactor (PWR) operated for approximately 40 years in Korea, was permanently ceased on June 18, 2017. To reduce worker exposure to radiation by reducing the dose rate in the system before starting main decommissioning activities, the permanently ceased Kori-1 NPP will be subjected to full system decontamination. Generally, the range of system decontamination includes Reactor Pressure Vessels (RPV), Pressurizer (PZR), Steam Generators (SG), Chemical & Volume Control System (CVCS), Residual Heat Removal System (RHRS), and Reactor Coolant System (RCS) piping. In order to decontaminate these systems and equipment in an effective manner, it is necessary to evaluate the influence of the flow characteristics in the RCS during the decontamination period. There are various methods of providing circulating flow rate to the system decontamination. In this paper, the flow characteristics in Kori-1 NPP reactor coolant according to RHR pump operation were evaluated. The evaluation results showed that system decontamination using an RHR pump was not effective at decontamination due first to impurities deposited in piping and equipment, and second to the extreme flow unbalance in the RCS caused deposition of impurities.