• Title/Summary/Keyword: Steam Generator(SG)

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Production of SCC Flaws and Evaluation Leak Behavior of Steam Generator Tubes (누설 및 파열실험용 SCC 결함 전열관 제작 및 누설거동 평가)

  • Hwang, Seong-Sik;Jung, Man-Kyo;Park, Jang-Yul;Kim, Hong-Pyo
    • Corrosion Science and Technology
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    • v.8 no.5
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    • pp.188-192
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    • 2009
  • A forced outage due to a steam generator tube leak in a Korean nuclear power plant was reported.1) Primary water stress corrosion cracking has occurred in many tubes in the plant, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to understand the leak behavior of the tubes containing stress corrosion cracks. Stress corrosion cracks were developed in 0.1 M sodium tetrathionate solution at room temperature. Steam generator(SG) tubes with short cracks were successfully fabricated with a restricted solution contact method. The leak rates of the degraded tubes were measured at room temperature. Some tubes with 100 % through wall cracks showed an increase of leak rate with time at a constant pressure.

EFFECTS OF AN ORIFICE-TYPE FLOW RESTRICTOR ON THE TRANSIENT THERMAL-HYDRAULIC RESPONSE OF THE SECONDARY SIDE OF A PWR STEAM GENERATOR TO A MAIN STEAM LINE BREAK (가압경수로 주증기관 파단시 증기발생기 2차측 과도 열수력 응답에 미치는 오리피스형 유량제한기의 영향)

  • Jo, J.C.;Min, B.K.
    • Journal of computational fluids engineering
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    • v.20 no.3
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    • pp.87-93
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    • 2015
  • In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

A Study on the Chemical Cleaning Process and Its Qualification Test by Eddy Current Testing

  • Shin, Ki Seok;Cheon, Keun Young;Nam, Min Woo;Min, Kyong Mahn
    • Journal of the Korean Society for Nondestructive Testing
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    • v.33 no.6
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    • pp.511-518
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    • 2013
  • Steam Generator (SG) tube, as a barrier isolating the primary coolant system from the secondary side of nuclear power plants (NPP), must maintain the structural integrity for the public safety and their efficient power generation. So, SG tubes are subject to the periodic examination and the repairs if needed so that any defective tubes are not in service. Recently, corrosion related degradations were detected in the tubes of the domestic OPR-1000 NPP, as a form of axially oriented outer diameter stress corrosion cracking (ODSCC). According to the studies on the factors causing the heat fouling as well as developing corrosion cracking, densely scaled deposits on the secondary side of the SG tubes are mainly known to be problematic causing the adverse impacts against the soundness of the SG tubes [1]. Therefore, the processes of various cleaning methods efficiently to dissolve and remove the deposits have been applied as well as it is imperative to maintain the structural integrity of the tubes after exposing to the cleaning agent. So qualification test (QT) should be carried out to assess the perfection of the chemical cleaning and QT is to apply the processes and to do ECT. In this paper, the chemical cleaning processes to dissolve and remove the scaled deposits are introduced and results of ECT on the artificial crack specimens to determine the effectiveness of those processes are represented.

A Study on Bagging Neural Network for Predicting Defect Size of Steam Generator Tube in Nuclear Power Plant (원전 증기발생기 세관 결함 크기 예측을 위한 Bagging 신경회로망에 관한 연구)

  • Kim, Kyung-Jin;Jo, Nam-Hoon
    • Journal of the Korean Society for Nondestructive Testing
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    • v.30 no.4
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    • pp.302-310
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    • 2010
  • In this paper, we studied Bagging neural network for predicting defect size of steam generator(SG) tube in nuclear power plant. Bagging is a method for creating an ensemble of estimator based on bootstrap sampling. For predicting defect size of SG tube, we first generated eddy current testing signals for 4 defect patterns of SG tube with various widths and depths. Then, we constructed single neural network(SNN) and Bagging neural network(BNN) to estimate width and depth of each defect. The estimation performance of SNN and BNN were measured by means of peak error. According to our experiment result, average peak error of SNN and BNN for estimating defect depth were 0.117 and 0.089mm, respectively. Also, in the case of estimating defect width, average peak error of SNN and BNN were 0.494 and 0.306mm, respectively. This shows that the estimation performance of BNN is superior to that of SNN.

Characteristic Analysis of Eddy Current Array Probe Signal in Combo Calibration Standard Tube Using Electromagnetic Numerical Analysis (전자기 수치해석을 이용한 표준보정시험편의 배열형 와전류 탐촉자 신호 특성 해석)

  • Kim, Ji-Ho;Lee, Hyang-Beom
    • Journal of the Korean Society for Nondestructive Testing
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    • v.30 no.4
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    • pp.330-337
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    • 2010
  • In this paper, 3-dimensional electromagnetic numerical analysis is performed about the eddy current(EC) array probe characteristic which is the next generation probe for accurate diagnosis of steam generator(SG) in nuclear power plants(NPPs). ASME(American Society of Mechanical Engineers) Standard and X-probe combo calibration standard tube are selected for acquisition of eddy current testing(ECT) signals and this result of compared with the real test signals for reasonability of result. Based on the analysis result of calibration standard tube, ECT signals that are about the defects of pitting, stress corrosion cracking(SCC), multiple SCC and wear is obtained. Material of specimen was Inconel 600 which is usually used for SG tubes in NPPs. The operation frequency of 300 kHz were used. The signal characteristics could be observed according to the various defects. The results in this paper can be helpful when the ECT signals from EC array probe are evaluated and analyzed.

Vibration Characteristics of Steam Generator U-tubes with Defect (결함을 가진 증기발생기 U-튜브의 진동특성)

  • 조종철;정명조;김웅식;김효정;김태형
    • Transactions of the Korean Society for Noise and Vibration Engineering
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    • v.13 no.5
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    • pp.400-408
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    • 2003
  • This paper investigates the vibration characteristics of steam generator (SG) U-tubes with defect. The operating SG shell-side flow field conditions for determining the fluidelastic instability parameters such as added mass are obtained from three-dimensional SG flow calculation. Modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, addressed is the effect of the internal pressure on the vibration characteristics of the tube.

WEAR BEHAVIOUR OF STEAM GENERATOR TUBES IN ROOM TEMPERATURE WATER

  • Lee, Young-Ho;Kim, Hyung-Kyu;Kim, In-Sup
    • Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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    • 2002.10b
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    • pp.203-204
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    • 2002
  • The wear behaviour of steam generator (SG) tubes (Inconel 600 and 690) against support materials (405 and 409 ferritic stainless steels) has been experimentally studied in room temperature water using reciprocating wear apparatus with tube-an-plate configuration. The results showed that the wear rate of Inconel 690 was lower than that of lnconel 600 with increasing normal loads and sliding amplitudes. Also, plastic deformation layers appear below the surface of both SG tubes, which have a specific thickness and are small compared with their grain size. This means that wear rate of SG tubes in water condition is closely related to the formation and fracture of plastic deformation layers.

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Fretting-Wear Characteristics of Steam Generator Tubes by Foreign Object

  • Jo Jong Chull;Jhung Myung Jo;Kim Woong Sik;Choi Young Hwan;Kim Hho Jung;Kim Tae Hyung
    • Nuclear Engineering and Technology
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    • v.35 no.5
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    • pp.442-453
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    • 2003
  • This study investigates the safety assessment of the potential for fretting-wear damages on steam generator (SG) U-tubes caused by foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions are obtained from three-dimensional SG flow calculation using the ATHOS3 code. Modal analyses are performed for the finite element modelings of U-tubes to get the natural frequency, corresponding mode shape and participation factor. The wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted. Also, discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube.

MIDLOOP Code Analysis of a ROSA-IV/LSTF Experiment for the Loss of Residual Heat Removal System Event During Mid- loop Operation

  • Han, Kee-Soo;Lee, Cheol-Sin;Park, Chul-Jin;Kim, Hee-Cheol
    • Proceedings of the Korean Nuclear Society Conference
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    • 1996.05b
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    • pp.683-690
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    • 1996
  • The MIDLOOP code has been developed for the evaluation of RES pressurization transients initiated from a loss-of-Residual Heat Removal System (RHRS) during mid-loop operation after reactor shutdown. It provides a fast running and realistic tool for studying parametrically the response of important plant parameters such as pressure, temperature, and level to various plant combinations of the primary side vent, makeup, and leakage procedures and the steam generator (SG) conditions. The code consists of ten nodes representing the primary and secondary sides of a nuclear power plant and can analyze the effect of air on the primary system pressurization and primary to secondary heat transfer. The analysis results of the MIDLOOP code are in good agreement with the ROSA-IV/LSTF experiment without opening in the RCS.

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Steady-State Performance Analysis of Pressurizer and Helical Steam Generator for SMART

  • Seo, Jae-Kwang;Kang, Hyung-Seok;Kim, Hwan-Yeol;Cho, Bong-Hyun;Lee, Doo-Jeong
    • Proceedings of the Korean Nuclear Society Conference
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    • 1997.05a
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    • pp.310-315
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    • 1997
  • System-Integrated Modular Advanced Reactor (SMART), where major primary components such as modular helical steam generator and self regulating pressurizer are integrated into reactor vessel, is currently under development. The pressurizer is designed to control the primary pressure mainly with partial pressure of nitrogen gas and to maintain the fluid temperature as low as possible for the purpose of minimizing steam contribution. The steam generator (SG) is designed to produce super-heated steam inside tube at power operation. Because the in-vessel pressurizer and in-vessel SG are classified as the characteristic components of SMART, it is important to perform a steady state calculation of these components in order to evaluate the adoption of these components. A steady state analysis of the in-vessel pressurizer and in-vessel SG has been performed under normal power operation and the results show an acceptable performance of the components.

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