• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.05b
• /
• pp.118-122
• /
• 1997

차세대 원자력발전소 증기발생기 전열관 재료로 채택된 니켈기저합금으로 기존 전열관 재료인 인코넬600에 비해 고온 고압 조건에서 응력부식균열에 강한 장점을 가진 합금인 인코넬690 시료에 최대 에너지 120 keV의 질소 이온빔을 조사하여 이 재료의 기계적 특성 변화를 관측하였다. 특성 시험으로는 표면 경화를 관찰하기 위한 미세 경도 시험을 수행하여 미세 경도 증가를 확인하였다 아울러 표면 경화가 피로 특성에 미치는 영향을 관찰하기 위해 피로 균열 전파 시험을 수행하여 이온 주입으로 인한 표면 경화가 피로 균열 전파를 촉진시킴을 관찰하였다.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.10a
• /
• pp.437-443
• /
• 1995

국내, 외에서 원자력발전소의 주요 구성 기기인 증기발생기의 세관 건전성과 관련 설계개선을 위한 연구가 활발히 진행되고 있다[2,3,4,5,6]. 현재 가동중인 발전소에서는 개선된 증기발생기로 교체하고자 하는 검토가 이루어지고 있으며, 설계중인 발전소에서는 중기발생기의 건전성을 향상시키기 위한 노력이 진행되고 있다. 본 논문에서 기존에 조사되고 검토된 자료를 바탕으로 [2] 현재까지 주로 사용되어온 증기발생기의 세관 재질을 인코넬 600 MA(mill annealed)로 사용할 때 40년 수명동안 증기발생기의 건전성을 보장할 수 있는 고온관의 온도를 분석한 결과 적절한 온도가 607$^{\circ}$F(319.4$^{\circ}C$)임을 알았다. 그리고 이 온도를 반영할 때 계통설계에 영향을 주는 설계사항에 대하여 검토하였고, 추가로 인코넬 600 MA보다 고온조건에서 건전성이 양호한 인코넬 690 TT(thermal treatment)를 사용할 때 설계에 미치는 영향도 검토하였다. 이러한 분석결과는 추후 국내 원자력발전소에서 보다 증기발생기의 건전성을 보장하기 위해 설계개선을 하고자 할 때 기초 자료가 되리라 판단한다.

• Proceedings of the KSME Conference
• /
• 2003.04a
• /
• pp.49-54
• /
• 2003

Most of the operating pressurized water reactors (PWRs)has chosen Inconel 600 as steam generator tubing. The long-term operation of steam generators showed that the use of this material induced localized corrosion damages. The current trend is using Inconel 690 as a tube material for the replacement steam generators. Based on the current trend, we have chosen Inconel 690 for the advanced Power Reactor 1400 (APR1400) steam generator tube material. In this paper, we examined the technical consideration in this modification: the effect of chemical composition, thermal conductivity, corrosion resistance and wear characteristics

• Proceedings of the KWS Conference
• /
• 1998.10a
• /
• pp.288-291
• /
• 1998

Inconel 600 alloy is used as the material of nuclear steam generator tubing because of its mechanical properties, formability, and corrosion properties. According to reports, the life time of nuclear power plants decreases because of the pitting, intergranular attack, primary water stress corrosion cracking(PWSCC), and intergranular stress corrosion cracking(IGSCC), and denting in the steam generator. The SCC test is very important because of SCC appears in various environment such as solutions, materials, and stress. The C-Rig specimen was made of the steam generator welded sleeve repairing by the pulsed Nd:YAG laser. In the corrosion invironment, corrosion solutions are Primary Water, Caustic, and Sulfate solution and corrosion time is 1624-4877hr. The permitted stress is 30-60ksi.In this C-Ring SCC test is the relationship between corrosion depth, crack and corrosion environment is evaluated. SCC was happens in Sulfate and Corrosion solution but doesn't happen in Primary Water. The corrosion time and stress is very affected by the severely environment of Sulfate or Caustic solution. The microstructure observation indicates that SCC causes interganular failure in the grain boundary of vertical direction.

• Journal of Welding and Joining
• /
• v.17 no.1
• /
• pp.82-87
• /
• 1999

The basic remote sleeve repair-welding technology by the pulsed Nd:YAG laser for increasing the lifetime of the steam generator tube in a nuclear power plant has been developed. The relationship between the connection width and welding parameters have been investigated for the fundamental research to apply the sleeve-repair-welding technique to the nuclear industry. The Inconel 600 tube and Inconel 690 sleeve used for high temperature and high pressure service were welded as round lap welding by Nd:YAG laser. It was observed that the tensile shear strength, 340MPa of the welded specimen is equivalent to about 60% of that of the base metal (Inconel 600), 550MPa. The difference between the hardness of the base metal and that of the laser welds was about 10%. Ductile fracture was partly occurred in the weld but the cracking has not been observed. In spite of absence of the crack, the strength of welds was not sufficient in terms of the tensile shear strength.

• Journal of Ocean Engineering and Technology
• /
• v.21 no.2 s.75
• /
• pp.50-59
• /
• 2007

Steam generator tubes are degraded from wear, stress corrosion cracking, rupture and fatigue and so on. Therefore, the failure assessment of steam generator tube is very important for the integrity of energy plants. In the steam generator tubes, sometimes, the local wall thinning may result from severe degradations such as erosion-corrosion damage and wear due to vibration. In this paper, the elasto-plastic analysis was performed by FE code ANSYS on steam generator tubes with wall thinning. Also, the four-point bending tests were performed on the wall thinned specimens, and then it was compared with the analysis results. We evaluated the failure mode, fracture strength and fracture behavior from the experiment and FE analysis. Also, it was possible to predict the crack initiation point by estimating true fracture ductility under multi-axial stress conditions at the center of the thinned area from FE analysis.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.14 no.1
• /
• pp.58-65
• /
• 2018

The purpose of this study is to develop a Half Nozzle Repair(HNR) process to prevent the leakage from welds on small bore piping in Reactor Coolant System. The Codes & Standards of tempered bead and design requirements of J-Groove welds are reviewed. Automatic machine GTAW welding and machining equipments are developed to perform HNR process. Single pass welding and overlay welding equipments are conducted in order to obtain the optimal temper bead welding process parameters with Alloy 52M filler wire. Coarse grain heat affected zone(CGHAZ) is formed by rapid cooling rate in heat affected zone after welding. Accordingly, a proper temper bead technique is required to reduce CGHAZ in 1-Layer of welds by 2- and 3-Layers. Mock-up tests show that the developed HNR process is possible to meet ASME Code & Standard requirements without any defect.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.36 no.6
• /
• pp.637-644
• /
• 2012

In a nuclear power plant, fretting wear due to impact motion between U-tubes and support structures located in steam generators can cause serious problems. In order to guarantee the reliability of the steam generator, the damage due to fretting wear should be thoroughly investigated. The purpose of this study is to elucidate the fretting wear mechanism qualitatively and quantitatively. Hence, fretting wear simulation is performed for the environments to which the actual steam generators in nuclear power plants are exposed. Initial experimental results are obtained for various experimental parameters, and the effect of the work rate and temperature on fretting wear is evaluated. In water, the wear coefficients for $90^{\circ}C$, $200^{\circ}C$, and $340^{\circ}C$ are found to be $9.051{\times}10^{-16}\;Pa^{-1}$, $3.009{\times}10^{-15}\;Pa^{-1}$, and $2.235{\times}10^{-15}\;Pa^{-1}$, respectively. It is also found that the wear coefficient at room temperature is larger than that at low temperature in water because of the dynamic viscosity of water.

• Composites Research
• /
• v.32 no.3
• /
• pp.158-162
• /
• 2019

Titanium carbide (TiC) reinforced Inconel 718 matrix composites were successfully fabricated by a novel liquid pressing infiltration process. Microstructure and mechanical properties of the fabricated 55 vol% TiC-Inconel 718 composite are analyzed. The composite exhibits superior mechanical properties, such as hardness and compressive strength as compared with Inconel 718. It is believed that Mo and Nb, which are alloying elements in the matrix, diffuse and solidify into the TiC reinforcement, resulting in generation of core-rim structure with excellent interfacial properties.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.28 no.4
• /
• pp.339-345
• /
• 2008

The heat exchanger tube in nuclear power plants is mainly fabricated from nonferromagnetic material such as a copper, titanium, and inconel alloy, but the moisture separator & reheater tube in the turbine system is fabricated from ferromagnetic material such as a carbon steel or ferrite stainless steel which has a good mechanical properties in harsh environments of high pressure and temperature. Especially, the moisture separator & reheater tubes, which use steam as a heat transfer media, typically employ a tubing with integral fins to furnish higher heat transfer rates. The ferromagnetic tube typically shows superior properties in high pressure and temperature environments than a nonferromagnetic material, but can make a trouble during the normal operation of power plants because the ferrous tube has service-induced damage forms including a steam cutting, erosion, mechanical wear, stress corrosion cracking, etc. Therefore, nondestructive examination is periodically performed to evaluate the tube integrity. Now, the remote field testing(RFT) technique is one of the solution for examination of ferromagnetic tube because the conventional eddy current technique typically can not be applied to ferromagnetic tube such as a ferrite stainless steel due to the high electrical permeability of ferrous tube. In this study, we have designed RFT probes, calibration standards, artificial flaw specimen, and probe pusher-puller necessary for field application, and have successfully carry out RFT examination of the moisture separator & reheater tube of nuclear power plants.