• Title, Summary, Keyword: Steam Generator Tube

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3-D Finite Element Analyses of Steam Generator Tubes Considering the Gap Effects (간극효과를 고려한 증기발생기 전열관의 3차원 유한요소해석)

  • Cho, Young Ki;Park, Jai Hak
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.7 no.4
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    • pp.51-56
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    • 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.

The Design Optimization of Preventive Measure Against APR1400 Steam Generator Tube Fretting Wear (신형경수로 증기발생기 마모손상 억제를 위한 설계최적화)

  • Lim, Hyuk-Soon;Park, Young-Sheop;Lee, Kwang-Han;Lee, Seok-Ho;Chung, Dae-Yul
    • Proceedings of the KSME Conference
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    • pp.2047-2052
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    • 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.

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Fluid-elastic Instability Evaluation of Steam Generator Tubes

  • Cho, Young Ki;Park, Jai Hak
    • International Journal of Safety
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    • v.11 no.1
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    • pp.1-5
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    • 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.

A Brief Review on the Design Factors of Steam Generator U-Tube Assembly for CANDU Type Nuclear Power Plant

  • Park, Nam-Il;Park, June-Soo
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.321-326
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    • 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.

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Best-Estimate Analysis of MSGTR Event in APR1400 Aiming to Examine the Effect of Affected Steam Generator Selection

  • Jeong, Ji-Hwan;Chang, Keun-Sun;Kim, Sang-Jae;Lee, Jae-Hun
    • Nuclear Engineering and Technology
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    • v.34 no.4
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    • pp.358-369
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    • 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.

Determination of Availability of Domestic Developed Bobbin Probe for Steam Generator Tube Inspection (증기발생기 전열관 와전류검사용 국내 개발 보빈탐촉자 적용성 분석)

  • Kim, In-Chul;Joo, Kyung-Mun;Moon, Yong-Sig
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.7 no.4
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    • pp.19-25
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    • 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.

Wear Progress Model by Impact Fretting in Steam Generator Tube (충격 프레팅에 의한 증기발생기 세관 마모손상 진행모델)

  • Lee, Jeong-Kun;Park, Chi-Yong;Kim, Tae-Ryong;Cho, Sun-Young
    • Proceedings of the KSME Conference
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    • pp.1684-1689
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    • 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.

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Wear Progress Model by Impact Fretting in Steam Generator Tube (충격 프레팅에 의한 증기발생기 세관 마모손상 진행모델)

  • Park, Chi-Yong;Lee, Jeong-Kun;Kim, Tae-Ryong
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.32 no.10
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    • pp.817-822
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    • 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.

Characteristics of Flow-induced Vibration for KSNP Steam Generator Tube at Concentrated Tube Plugging Zone (한국표준원전 증기발생기의 관막음 집중 영역 근방에서의 유체유발진동 특성해석)

  • 유기완;조봉호;박치용;박수기
    • Transactions of the Korean Society for Noise and Vibration Engineering
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    • v.13 no.6
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    • pp.452-459
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    • 2003
  • The characteristics of fluid-elastic instability and effects of turbulent excitations for the KSNP steam generator tubes were investigated numerically. The information for the thermal-hydraulic data of the steam generator has been obtained by using the ATHOS3-MOD1 code and the flow-induced vibration(FIV) analysis has been conducted by using the PIAT(program for Integrity assessment of SG tube) code. The KSNP steam generator has the concentrated plugging zone at the vicinity of the stay cylinder inside the SG. To investigate the cause of the concentrated tube plugging zone, the FIV analysis has been performed for various column and row number of the steam generator tubes. From the results of FIV analysis the stability ratio due to the fluid-elastic instability and vibrational amplitude due to the turbulent excitation in the concentrated plugged zone have a trend of larger values than those of the outer concentrated tube Plugging zone.

Wolsong Unit 1 Steam Generator Aging Management for Continued Operation (월성 1호기 계속운전을 위한 증기발생기 열화관리)

  • Song, Myung Ho;Kim, Hong Key;Lee, Jung Min
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.6 no.2
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    • pp.28-33
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    • 2010
  • As a part of license renewal for the continued operation of Wolsong unit 1, the periodic safety review report was submitted near the end of design lifetime, 2012, and now is under reviewing. Major components of primary system such as pressure tubes, feeder pipes and so on are being replaced and many components of secondary system are also being repaired. So the license renewal of Wolsong unit 1 is expected to be acquired without significant issues. But on the other hand, steam generators of Wolsong unit 1 had the good performance and therefore the replacement and repair for the steam generator are not needed. Recently it is reported that some cracks were detected in a few of european steam generator with Alloy 800 tubes and the cause of cracks was the outer diameter stress corrosion cracking(ODSCC) due to the concentration of chemical impurities on the outer surface of tube. Accordingly the overall review on this issue was performed. The long-term operation is likely to results to form the concentration mechanism for the tube corrosion as the sludge build-up in the secondary side of steam generator and the crack in the crevice between tube and tube-sheet and expansion transitions is apt to be occurred. In this paper, the history of steam generator inspection and operation of Wolsong unit 1 are reviewed and the reliability of steam generator tube is evaluated and the steam generator aging management program for Wolsong unit 1 is introduced.

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