• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2004년도 춘계학술대회
• /
• pp.2047-2052
• /
• 2004

Inconel-600 alloy has been used as steam generator tube material for current pressurized water reactors (PWRs). The long-term operation of steam generators showed that the use of this material induced localized corrosion damages and increased tube wear of steam generator. To protect these problems, steam generator tube material is being changed to Inconel-690 alloy. Based on the current trend, we have chosen Inconel 690 as the Advanced Power Reactor 1400 (APR1400) steam generator(SG) tube material and performed the design optimization of preventive measure against tube fretting wear for the APR1400 steam generator. In this paper, we examined the technical consideration in this modification : the selection of material, wear characteristics, effect of the Egg-crate Flow Distribution Plate installation, and effect analysis of vertical strip installation.

• 한국압력기기공학회 논문집
• /
• 제7권4호
• /
• pp.51-56
• /
• 2011

Steam generator is one of the main equipments that affect safety and long term operation in nuclear power plants. Fluid flows inside and outside of the steam generator tubes and induces vibration. To prevent the vibration the tubes are supported by AVB (anti vibration bar). When the steam generator tube contact to AVB, it is damaged by the accumulation of wear and corrosion. Therefore studies are required to determine the effects of the gap between the steam generator tube and AVB. In order to obtain the stress and the displacement distributions of the steam generator tube, three dimensional finite element analyses were performed by using the commercial program ANSYS. Using the calculated the stress and the displacement distributions, the static residual strength of the steam generator tube can be evaluated. The results show that the stress and displacement of the steam generator tube increase significantly compared with the results from a zero-gap model.

• Nuclear Engineering and Technology
• /
• 제54권7호
• /
• pp.2702-2713
• /
• 2022

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release. AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences. The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number. The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.

• International Journal of Safety
• /
• 제11권1호
• /
• pp.1-5
• /
• 2012

It has been reported that the plugged steam generator tube of Three Mile Island Unit 1 in America was damaged by growing flaw and then this steam generator tube destroyed the nearby steam generator tubes of normal state. On this account, stabilizer installation is necessary to prevent secondary damage of the steam generator tubes. The flow-induced vibration is one of the major causes of the fluid-elastic instability. To guarantee the structural integrity of steam generator tubes, the flow-induced vibration caused by the fluid-elastic instability is necessary to be suppressed. In this paper, the effective velocity and the critical velocity are calculated to evaluate the fluid-elastic instability. In addition, stability ratio value of the steam generator tubes is evaluated in order to propose one criterion when to determine stabilizer installation.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(4)
• /
• pp.321-326
• /
• 1996

During the plant operation, steam generator U-tube assembly will potentially be subject to adverse environmental conditions which can cause damages to them. This report addresses the major design factors of CANDU type steam generator which are intended to minimize the potential tube damages. Such factors include U-tube material, high circulation ratio, tube-to-tubesheet joint, tube support design. Also a few suggestions are presented for the design and performance improvement of CANDU type steam generators.

• Nuclear Engineering and Technology
• /
• 제34권4호
• /
• pp.358-369
• /
• 2002

Abundant information about analyses of single steam generator tube rupture (SGTR) events is available because of its importance in terms of safety. However, there are few literatures available on analyses of multiple steam generator tube rupture (MSGTR) events. In addition, knowledge of transients and consequences following a MSGTR event are very limited as there has been no occurrence of MSGTR event in the commercial operation of nuclear reactors. In this study, a postulated MSGTR event in an APR1400 is analyzed using thermal-hydraulic system code MARSI.4. The present study aims to examine the effects of affected steam generator selection. The main steam safety valve (MSSV) lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generate. (S/G) is affected. The comparison shows that the cases where two steam generators are simultaneously affected allow longer time for operator action compared with the cases where a single steam generator is affected. Furthermore, the tube ruptures in the steam generator where a pressurizer is connected leads to the shortest operator response time.

• 한국압력기기공학회 논문집
• /
• 제7권4호
• /
• pp.19-25
• /
• 2011

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which is the pressure boundary between the primary and secondary systems. The integrity of SG tube has been confirmed by the eddy current test every outage. The eddy current technique adopting bobbin probe is currently the primary technique for the steam generator tubing integrity assesment. The bobbin probe is one of the essential components which consist of the whole ECT examination system and provides us a decisive data for the evaluation of tube integrity. Until now, all of the ECT bobbin probes in Korea which is necessary to carry out inspection are imported from overseas. However, KHNP has recently developed the bobbin probe design technology and transferred it to domestic manufacturers to fabricate the probes. This study has been conducted to establish technical requirements applicable to the steam generator tube inspection using the bobbin probes fabricated by the domestic manufactures. The results have been compared with the results obtained by using foreign probe to identify the availability to the steam generator tube inspection. As a result, it is confirmed that the domestic bobbin probe is generally applicable to SG tube inspection in the NPPs.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2007년도 춘계학술대회A
• /
• pp.1684-1689
• /
• 2007

Fretting wear is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Especially, impact fretting wear occurred between steam generator tubes and tube support plates or anti-vibration bar. Various tests have been carried out to investigate the wear mechanisms and to report the wear coefficients. Those are fruitful to get insight for the wear damage of steam generator tubes; however, most wear researches have concentrated on sliding wear of the steam generator tubes, which may not represent the wear loading modes in real plants. In the present work, impact fretting tests of steam generator tube were carried out. A wear progression model for impact-fretting wear has been investigated and proposed. The proposed wear progression model of impact-fretting wear is as follows; oxide film breaking step at the initial stage, and layer formation step, energy accumulation step and finally particle torn out step which is followed by layer formation in the stable impact-fretting progress. The wear coefficient according to the work-rate model has been also compared with one between tube and support.

• 대한기계학회논문집A
• /
• 제32권10호
• /
• pp.817-822
• /
• 2008

Fretting wear is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Especially, impact fretting wear occurred between steam generator tubes and tube support plates or anti-vibration bar. Various tests have been carried out to investigate the wear mechanisms and to report the wear coefficients. Those are fruitful to get insight for the wear damage of steam generator tubes; however, most wear researches have concentrated on sliding wear of the steam generator tubes, which may not represent the wear loading modes in real plants. In the present work, impact fretting tests of steam generator tube were carried out. A wear progress model for impact-fretting wear has been investigated and proposed. The proposed wear progress model of impact-fretting wear is as follows; oxide film breaking step at the initial stage, and layer formation step, energy accumulation step and finally particle torn out step which is followed by layer formation in the stable impact-fretting progress. The wear coefficient according to the work-rate model has been also compared with one between tube and support.

• 한국윤활학회:학술대회논문집
• /
• 한국윤활학회 2001년도 제34회 추계학술대회 개최
• /
• pp.59-62
• /
• 2001

Flow induced vibration in steam generators has caused dynamic interactions between tubes and contacting materials resulting in fretting wear . Series of experiments have been performed to examine the wear properties of Inconel 690 steam generator tubes in various environmental conditions. For the present study, the test rig was designed to examine the fretting wear and rolling wear properties in high temperature(room temperature - 290。C) water. The test was performed at constant applied load and sliding distance to investigate the effect of test temperature on wear properties of the steam generator tube materials. To investigate the wear mechanism of material, the worn was observed using scanning electron microscopy. The weight loss increase at higher test temperature was caused by the decrease of water viscosity and the mechanical property change of tube material. The mechanical property changes of steam generator tube material, such as decrease of hardness or yield stress in the high temperature tests. From the SEM observation of worn surfaces, the severe wear scars were observed in specimens tested at the higher temperature.