• 제목/요약/키워드: Steam generator (SG)

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Structural Integrity Evaluation of SG Tube with Surface Wear-type Defects (표면 마모결함을 고려한 증기발생기 세관의 구조건전성 평가)

  • Kim, Jong-Min;Huh, Nam-Su;Chang, Yoon-Suk;Hwang, Seong-Sik;Kim, Joung-Soo;Kim, Young-Jin
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.30 no.12 s.255
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    • pp.1618-1625
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    • 2006
  • During the last two decades, several guidelines have been developed and used for assessing the integrity of a defective steam generator (SG) tube that is generally caused by stress corrosion cracking or wall-thinning phenomenon. However, as some of SG tubes are also failed due to fretting and so on, alternative failure estimation schemes are required for relevant defects. In this paper, parametric three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of SG tubes with different defect configurations; elliptical wear, tapered and flat wear type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of SG tube. After investigating the effect of key parameters such as defect depth, defect length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wear region. Comparison of failure pressures predicted by the proposed estimation scheme with corresponding burst test data showed a good agreement.

OPTIMIZATION OF THE PARAMETERS OF FEEDWATER CONTROL SYSTEM FOR OPR1000 NUCLEAR POWER PLANTS

  • Kim, Ung-Soo;Song, In-Ho;Sohn, Jong-Joo;Kim, Eun-Kee
    • Nuclear Engineering and Technology
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    • v.42 no.4
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    • pp.460-467
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    • 2010
  • In this study, the parameters of the feedwater control system (FWCS) of the OPR1000 type nuclear power plant (NPP) are optimized by response surface methodology (RSM) in order to acquire better level control performance from the FWCS. The objective of the optimization is to minimize the steam generator (SG) water level deviation from the reference level during transients. The objective functions for this optimization are relationships between the SG level deviation and the parameters of the FWCS. However, in this case of FWCS parameter optimization, the objective functions are not available in the form of analytic equations and the responses (the SG level at plant transients) to inputs (FWCS parameters) can be evaluated by computer simulations only. Classical optimization methods cannot be used because the objective function value cannot be calculated directely. Therefore, the simulation optimization methodology is used and the RSM is adopted as the simulation optimization algorithm. Objective functions are evaluated with several typical transients in NPPs using a system simulation computer code that has been utilized for the system performance analysis of actual NPPs. The results show that the optimized parameters have better SG level control performance. The degree of the SG level deviation from the reference level during transients is minimized and consequently the control performance of the FWCS is remarkably improved.

Investigation on reverse flow characteristics in U-tubes under two-phase natural circulation

  • Chu, Xi;Li, Mingrui;Chen, Wenzhen;Hao, Jianli
    • Nuclear Engineering and Technology
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    • v.52 no.5
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    • pp.889-896
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    • 2020
  • The vertically inverted U-tube steam generator (UTSG) is widely used in the pressurized water reactor (PWR). The reverse flow behavior generally exists in some U-tubes of a steam generator (SG) under both single- and two-phase natural circulations (NCs). The behavior increases the flow resistance in the primary loop and reduces the heat transfer in the SG. As a consequence, the NC ability as well as the inherent safety of nuclear reactors is faced with severe challenges. The theoretical models for calculating single- and two-phase flow pressure drops in U-tubes are developed and validated in this paper. The two-phase reverse flow characteristics in two types of SGs are investigated base on the theoretical models, and the effects of the U-tube height, bending radius, inlet steam quality and primary side pressure on the behavior are analyzed. The conclusions may provide some promising references for SG optimization to reduce the disadvantageous behavior. It is also of significance to improve the NC ability and ensure the PWR safety during some accidents.

Motion planning of a steam generator mobile tube-inspection robot

  • Xu, Biying;Li, Ge;Zhang, Kuan;Cai, Hegao;Zhao, Jie;Fan, Jizhuang
    • Nuclear Engineering and Technology
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    • v.54 no.4
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    • pp.1374-1381
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    • 2022
  • Under the influence of nuclear radiation, the reliability of steam generators (SGs) is an important factor in the efficiency and safety of nuclear power plant (NPP) reactors. Motion planning that remotely manipulates an SG mobile tube-inspection robot to inspect SG heat transfer tubes is the mainstream trend of NPP robot development. To achieve motion planning, conditional traversal is usually used for base position optimization, and then the A* algorithm is used for path planning. However, the proposed approach requires considerable processing time and has a single expansion during path planning and plan paths with many turns, which decreases the working speed of the robot. Therefore, to reduce the calculation time and improve the efficiency of motion planning, modifications such as the matrix method, improved parent node, turning cost, and improved expanded node were proposed in this study. We also present a comprehensive evaluation index to evaluate the performance of the improved algorithm. We validated the efficiency of the proposed method by planning on a tube sheet with square-type tube arrays and experimenting with Model SG.

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.

Simulation of Water/steam into Sodium Leak Behavior for an Acoustic Noise Generation Mechanism Study

  • Kim, Tae-Joon;Hwang, Sung-Tai;Jeong, Kyung-Chai;Park, Jong-Hyeun;Valery S. Yughay
    • Nuclear Engineering and Technology
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    • v.33 no.2
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    • pp.145-155
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    • 2001
  • This simulation first allows us to define a transition zone from a bubble to jet mode of the argon out-flow and hereinafter to define a similar area for water-steam leak in the KALIMER SG (Korea Advanced Liquid Metal Reactor Steam Generator) using a water mock-up system, taking into account the KALIMER leak classification and tube bundle design, as a simulation of a real water-steam into sodium leak. in accordance with leak conditions in the KALIMER SG, the transition from bubbling to jetting is studied by means of turbulence regime simulation for argon out-flow through a very small orifice, which has the equivalent diameter of about 0.253 mm. finally the noise generation mechanism is explained from the existing experimental data. We also confirmed the possibility of micro-leak detection from the information of the bubbling mode through simulations and the experiment in this study.

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Key Findings from the Artist Project on Aerosol Retention in a Dry Steam Generator

  • Dehbi, Abdelouahab;Suckow, Detlef;Lind, Terttaliisa;Guentay, Salih;Danner, Steffen;Mukin, Roman
    • Nuclear Engineering and Technology
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    • v.48 no.4
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    • pp.870-880
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    • 2016
  • A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

Signal Analysis of Eddy Current Array Probe According to Size Variation of FBH Defects (배열 와전류 프로브의 FBH 결함 크기 변화에 따른 신호 해석)

  • Kim, Ji-Ho;Lim, Geon-Gyu;Lee, Hyang-Beom
    • Journal of the Korean Society for Nondestructive Testing
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    • v.29 no.2
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    • pp.137-144
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    • 2009
  • In this paper, the signal analysis of eddy current array probe was performed to analyze the electromagnetic characteristics with the variation of FBH(flat bottomed hole) defects size on steam generator tube in NPP(nuclear power plants) using the electromagnetic finite element method. To obtain the electromagnetic characteristic of probes, the governing equation was derived from Maxwell's equations, and the individual problem was analyzed by using the 3-dimensional finite element method. For the simulation FBH defects were used. The depth of FBH defects were 20%, 40%, 60%, 80% and 100% of steam generator(SG) tube thickness, and it was assumed that the defects were located on the tube outside. And the operation frequencies of 100 kHz, 300 kHz and 400 kHz were used. Material of specimen was Inconel 600 which is usually used for SG tubes in NPP. The signal difference could be observed according to the size variation of depth of FBH defects and operation frequencies. The results in this paper can be helpful when the ECT(eddy current testing) signals from EC array probe are evaluated and analyzed.

Analysis of Fluid-elastic Instability In the CE-type Steam Generator Tube (CE형 증기발생기 전열관에 대한 유체탄성 불안정성 해석)

  • 박치용;유기완
    • Transactions of the Korean Society for Noise and Vibration Engineering
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    • v.12 no.4
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    • pp.261-271
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    • 2002
  • The fluid-elastic instability analysis of the U-tube bundle inside the steam generator is very important not only for detailed design stage of the SG but also for the change of operating condition of the nuclear powerplant. However the calculation procedure for the fluid-elastic instability was so complicated that the consolidated computer program has not been developed until now. In this study, the numerical calculation procedure and the computer program to obtain the stability ratio were developed. The thermal-hydraulic data in the region of secondary side of steam generator was obtained from executing the ATHOS3 code. The distribution of the fluid density can be calculated by using the void fraction, enthalpy, and operating pressure. The effective mass distribution along the U-tube was required to calculate natural frequency and dynamic mode shape using the ANSYS ver. 5.6 code. Finally, stability ratios for selected tubes of the CE type steam generator were computed. We considered the YGN 3.4 nuclear powerplant as the model plant, and stability ratios were investigated at the flow exit region of the U-tube. From our results, stability ratios at the central and the outside region of the tube bundle are much higher than those of other region.

Degradation Characteristics of Tubes in the Steam Generator Tubesheet (증기발생기 관판내부 균열 열화 특성)

  • Cho, Nam Cheoul;Kang, Yong Suk;Kim, Heung Nam;Lee, Kuk-Hee
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.10 no.1
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    • pp.7-14
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    • 2014
  • There has been extensive experience associated with the operation of SGs wherein it was believed, based on NDE, that throughwall tube indications were present within the tubesheet. The installation of the SG tubes usually involves the development of a short interference fit, referred to as the tack expansion, at the bottom of the tubesheet. The tack expansion was usually effected by a hard rolling process and thereafter, in most instance, by the expansion of a urethane plug inserted into the tube end and compressed in the axial direction. The rolling process by its very nature is considered to be intensive with regard to metalworking at the inside surface of the tube and would be expected to lead to higher residual surface stresses. Alternate repair criteria(ARC) in the tack expansion area have been developed and applied to nuclear power plants in USA, however domestic nuclear power plants have not applied ARC for tubes in tubeheet area yet. In consideration of the degradation characteristics of tubes in the Steam Generator tubesheet, this paper suggests ARC application for tubes in the steam generator tubesheet of the domestic nuclear power plants in order to assure life time of the steam generator as well as nuclear power plants.