• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1847-1861
• /
• 2020

A design study was performed to improve the limit aseismic performance (LSP) of a primary heat transport system (PHTS) pump. This pump is part of the primary equipment of a prototype generation IV sodium-cooled fast reactor (PGSFR). The LSP is the maximum allowable seismic load that still ensures structural integrity. To calculate the LSP of the PHTS pump, a structural analysis model of the pump was developed and its dynamic characteristics were obtained by modal analysis. The floor response spectrum (FRS) initiated from a safety shutdown earthquake (SSE), 0.3 g, was applied to the support points of the PHTS pump, and then the seismic induced stresses were calculated. The structural integrity was evaluated according to the ASME code, and the LSP of the PHTS pump was calculated from the evaluation results. Based on the results of the modal analysis and LSP of the PHTS pump, design parameters affecting the LSP were selected. Then, ways to improve the LSP were proposed from sensitivity analysis of the selected design variables.

• Proceedings of the Korean Society for Technology of Plasticity Conference
• /
• 2006.05a
• /
• pp.136-140
• /
• 2006

We are developing a vol-oxidizer which transforms the spent $UO_2$ pellets into the $U_3O_8$ power through oxidizing process. The vol-oxidizer consists of furnace, filter, heater and valve etc. When the filter is blocked by the powder, the internal pressure of the furnace is increased owing to the air flow restriction. Then, the furnace vessel is swelled and deformed by it. In this paper, we proposed a procedure of the thermal analysis for furnace vessel design of spent fuel vol-oxidizer. In this work, we determined the thickness of the furnace through analyzing the internal pressure and the thermal stress of the furnace with respect to various pressure and temperature. To analyze the thermal stress, we used ANSYS 8.0 for constructing a FEM model of the furnace, and then analyzed it based on the ASME code. We also surveyed the material property and yield stress of SUS304 with various temperature. Analysis results are compared with the yield stress of the material. We manufactured the furnace and conduct the verification experiments.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.26 no.7
• /
• pp.1376-1383
• /
• 2002

Seismic qualification of the Air Cleaning Units for nuclear power plant Ulchin 5&6 has been performed with the guideline of ASME Section III and IEEE 344 code. By using the structural and geometrical similarity analysis, the three models to be analyzed are condensed into a single model and, at the same time, the excitation forces and other operating loads for each model are encompassed with respect to different loading conditions. As the fundamental frequencies of the structure are found to be less than 33Hz, which is the upper frequency limit of the seismic load, response spectrum analysis using ANSYS is performed in order to combine the modal stresses within the frequency limit. In order to confirm the structural and electric stability of the major components, modal analysis theory is adopted to derive the required response spectrum at the component locations. As the all combined stresses obtained from the above procedures are less than allowable stresses and no mechanical or electrical failures are found from the seismic testing, the authors confirm the safety of the nuclear equipments Air Cleaning Units studied in this paper.

• Journal of KSNVE
• /
• v.9 no.3
• /
• pp.535-543
• /
• 1999

The seismic qualification of the structures has been great concern in our engineering society with an effort to reduce the severe damages from an earthquake. However, on the contrary to the importance of the seismic qualification, the whole procedures are used to rely on the advanced countries who require various expenses for the qualification, which leads to the heavy loss of the foreign currency. In this study, the nuclear air conditioning system produced by LG Cable are adopted for the seismic qualification based on the guideline of NUREG, IEEE and ASME code. In order to confirm the validity of the present study, the results from the Ellis & Watts are compared with the present results and, also, the seismic qualification procedures and results mentioned herein are approved by KOPEC, which is a naitonal surveillance institute for the construction of nuclear power plant. From these results, the author confirmed the validity of the present seismic qualification procedures and results, which might be usefully applied to the other kind of seismic qualification of equipments.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.29 no.8 s.239
• /
• pp.1070-1077
• /
• 2005

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 collinear or parallel two adjacent axial through-wall cracks. The objective of this study is to examine the failure prediction models and propose optimum ones for non-aligned two 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 clacks. Finally, the optimum failure prediction model was used to demonstrate the conservatism of flaw characterization rules for various multiple flaws according to ASME code.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.26 no.1
• /
• pp.179-187
• /
• 2002

Finite element analysis was performed to evaluate the integrity of the tubes in the fixed tubesheet of horizontal type heat exchanger under operating condition. For the finite element analysis of the heat exchanger, tubes and tubesheets were equivalently modeled with concentroidal hexagonal columns and solid plates having equivalent properties for the convenience of finite element modeling, respectively. Load combination of tube pressure and thermal expansion most likely to precipitate possible failure of the tubes was selected and applied to the finite element analysis. The compressive stresses of the tubes were calculated based on displacements of each tube, which were obtained from anile element analysis. Finally, the maximum tube stress was compared with the design criterion of ASME Boiler and Pressure Vessel Code Section VIII.

• Proceedings of the KSME Conference
• /
• 2000.11a
• /
• pp.212-218
• /
• 2000

The importance and role of pre-/in-service inspection(PSI/ISI) for nuclear power plant(NPP) components are intimately related to plant design, safety, reliability, operation, etc. In this paper, for an effective and efficient management of large amounts of PSI/ISI data in NPPs, an intelligent database program(WS-IDPIN) for PSI/ISI data management of NPP was developed. WS-IDPIN program enables the prompt extraction of previously conducted PSI/ISI conditions and results so that the time-consuming data management, painstaking data processing and analysis in the past are avoided. Furthermore, development of ISI UT auto flaw evaluation and acceptance module based on ASME Code Sec. XI were presented. This module can be used for any angle beam examination from flat plate to spherical shapes as selected by the proper azimuthal angle. This program can be further developed as a unique PSI/ISI data management expert system.

• Journal of the Korean Society of Manufacturing Process Engineers
• /
• v.17 no.4
• /
• pp.23-31
• /
• 2018

Stress of a steam separator, equipment of the high-pressure (HP) evaporator for a HRSG, was analyzed and evaluated according to ASME Boiler & Pressure Vessel Code Section VIII Division 2. First, from the analysis results of the piping system model of the HP evaporator, reaction forces of the riser tubes connected to the steam separator, i.e., nozzle loads, were derived. Next, a finite element model of the steam separator was constructed and analyzed for the design pressure and the nozzle loads. The results show that the maximum stress occurred at the bore of the riser nozzle. The primary membrane stresses at the shell and nozzle were found to be less than the allowable stress. Next, the steam separator was analyzed for the steady-state operating conditions of operating pressure, operating temperature, and nozzle loads. The maximum stress occurred at the bore of the riser nozzle. The primary plus secondary membrane plus bending stress at the shell and nozzle was found to be less than the allowable stress.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.11 no.2
• /
• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.11 no.2
• /
• pp.1-5
• /
• 2015

A failure or degradation of reactor upper head penetration is a troublesome problem at Nuclear Power Plants. A flaw in the reactor upper head penetration can result in unplanned plant shutdown for repair, and cause serious economic losses on the plants. Consequently, a detection of flaws is a matter of more importance. Until now, only the base metal, not including J-groove weld, in reactor upper head penetration has been inspected in accordance with 10 CFR 50.55a and ASME code case N-729-1 requirements. Accordingly, it is rather difficult to detect manufacturing defects and repair defects in J-groove weld. This paper presents a case study on the application of Time of Flight Diffraction UT technique to examine the J-groove weld in reactor head penetration using reactor head penetration mockup with artificial flaws. We expect that this study result will offer a way to understand the non-destructive examination technology for J-groove weld in reactor upper head penetration.