• 한국정밀공학회지
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• 제16권12호
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• pp.78-86
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• 1999

The improvement of the water level control is important since it will prevent the steam generator trip so that improve the reliability and credibility of operation system. In this paper, the closed loop system identification is performed which can be used for the system monitoring and prediction of the system response. The model also can be used for the prediction control. Irving model is used as a steam generator model. The plant is an open loop unstable and non-minimum phase system. Fuzzy controller stabilize the system and the stable controller stabilize the system and the stable closed loop system is identified using neural networks. The obtained neural network model is validated using the untrained input and output. The results of computer simulation show the obtained Neural Network model represents the closed loop system well.

• 한국재료학회지
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• 제11권10호
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• pp.841-845
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• 2001

The steam generator tubes of power plant damaged by sliding wear due to flow-induced motion of foreign object. Amount of wear have been predicted by Achard's wear equation until now. However, there are large error and low reliability, because this equation regards wear coefficient(k) as constant. The sliding wears tests have been performed at room temperature to examine parameters of wear (wear distance, contact stress). The steam generator tube material for wear test is used Inconel 600 and foreign object material is used 304 austenite stainless steel. The sliding wear tests show that the amount of wear is not linearly proportional to the wear distance(for 374 austenite stainless steel). According to experimental result, wear coefficient is not constant k but function k(s) of wear distance. The newly modified wear predictive equation V=k(s)F have small error and high reliability.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2000년도 하계학술대회 논문집 D
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• pp.2578-2580
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• 2000

A siumlator had been developed and used for reliability verification on medium size steam turbine control programs prior to its actual operation in field. A mathematical model on thermal dynamics pertaining to prime mover steam turbine and electrical generator was realized and included in this simulator. Also, many operating data acquired from fields was utilized in order to decide mechanical and thermal dynamic characteristics such as friction loss, windage loss and inertia. A user can decide closing or opening velocity of steam stop valve and steam regulation valve. This simulator is able to generate steam pressure, turbine speed, electrical power, and power system frequency.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2001년도 하계학술대회 논문집 D
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• pp.2138-2140
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• 2001

A siumlator had been developed and will be used for reliability verification on large steam turbine control programs prior to its actual operation in field. A mathematical model on thermal dynamics pertaining to prime mover steam turbine and electrical generator was realized and included in this simulator. Also, many operating data acquired from fields was utilized in order to decide mechanical and thermal dynamic characteristics such as friction loss, windage loss and inertia. A user can decide closing or opening velocity of steam stop valves and steam regulation valves. This simulator is able to generate steam pressure, turbine speed, electrical power, and power system frequency.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1991년도 하계학술대회 논문집
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• pp.730-733
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• 1991

The inspection for steam generator tubes of nuclear power plants is performed by eddy current test method. In the current, human experts should check enormous amounts of eddy current(EC) signals to find abnormal ones on the computer screen. This method could cause a few problems. The purpose of this paper is to develop an expert system which can automatically evaluate EC signals of steam generator tubes. Since this expert system can replace or help human experts, the reliability in EC signal evaluation can be improved, and the required man-power can be reduced. Additionally, application of this system can shorten the overhaul period, contribute to a safe operation of the nuclear power plant.

• Nuclear Engineering and Technology
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• 제36권4호
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• pp.327-337
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• 2004

It is commonly required that tubes with defects exceeding $40\%$ of wall thickness in depth should be plugged; however, this criterion is too conservative for some locations and for some types of defects. Many studies have been done with the aim of developing an alternative plugging criteria, and these studies have shown that steam generator tubes with a certain range of axial through-wall cracks could remain in service without any safety or reliability problems. However, these studies have been limited, thus far, to consideration of single cracked tubes, necessitating a study on multiple cracks, which are commonly found. A crack coalescence model applicable to steam generator tubes with two collinear axial through-wall cracks was proposed in the previous study. In this paper, the investigation is extended to the parallel axial cracks spaced in a circumferential direction, because parallel axial cracks are more frequently detected during in-service inspections than collinear axial cracks. Interaction effects between two parallel cracks are evaluated by performing elastic and elastic-plastic finite element analyses.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1381-1389
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• 2019

Stress corrosion cracking (SCC) widely found in both primary and secondary sides of steam generator (SG) tubing in pressurized water reactors (PWR) has become an important safety issue. Using eddy-current tests (ECTs), non-destructive evaluations are performed for the integrity management of SG tubes against intergranular SCC. To enhance the reliability of ECT, this study investigates the effects of oxide films on ECT's detection capabilities for SCC in laboratory-degraded SG tubing in high temperature and high pressure aqueous environment.

• 한국안전학회지
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• 제35권4호
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• 비파괴검사학회지
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• 제21권5호
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• pp.501-509
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• 2001

증기발생기 전열관에서 1차측 응력부식균열의 발생빈도가 증가하고 있으므로 이의 정확한 탐지와 평가를 위해서는 균열 형상에 따른 와전류 신호특성을 규명하고 적합한 탐촉자를 선정하는 것이 매우 중요하다. 본 연구에서는 증기발생기 전열관의 내면 축방향 균열에 대한 와전류 검사의 검출능과 크기예측에 대한 신뢰도를 정량적으로 평가하고 pancake coil과 plus coil과의 신호특성 차이를 비교하였다. 이를 위하여 전열관 내면에 EDM으로 노치를 가공한 시편과 실제 증기발생기에서 1차측 응력부식균열이 발생하여 인출한 전열관을 시험편으로 사용하였다. 본 연구에서 얻어진 결과를 토대로 내면 축방향 균열에 대한 와전류 검사 신뢰도 향상을 위한 방안을 제시하였다.

• International Journal of Safety
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• 제3권1호
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• pp.53-58
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• 2004

SMART-P is a promising advanced small and medium category nuclear power reactor. It is an integral type reactor with a sensible mixture of new innovative design features and proven technologies aimed at achieving a highly enhanced safety and improved economics. The enhancement of the safety and reliability is realized by incorporating inherent safety improving features and reliable passive safety systems. The improvement in the economics is achieved through a system simplification, and component modularization. Preliminary safety analyses on selected limiting accidents confirm that the inherent safety improving design characteristics and the safety system of SMART-P ensure the reactor's safety. SMART-P is an advanced integral pressurized water reactor. The purpose of this study is for the safety analysis of the steam generator module pipe break for the SMART-P. The integrity of the fuel rod is the major criteria of this analysis. As a result of this analysis, the safety of the RCS and the secondary system is guaranteed against the module pipe break of a steam generator of the SMART-P.