• 대한기계학회논문집A
• /
• 제35권8호
• /
• pp.951-956
• /
• 2011

원자력 발전설비의 증기발생기용 세관 재료로 인코넬 합금이 사용되고 있다. 인코넬 합금은 고니켈과 크롬의 합금으로 고온에서 내부식성이 뛰어난 특성을 가지고 있다. 본 연구에서는 인코넬 600 과 690 합금에 대해 실제 원전의 운전온도, 즉 $320^{\circ}C$가 프레팅 피로 거동에 미치는 영향에 대해 연구해보았다. 그 결과 $320^{\circ}C$에서의 단순 피로한도와 프레팅 피로한도가 상온에서의 단순 및 프레팅 피로한도에 비해 다소 저하하는 것을 알 수 있었다. 아울러 하중 반복회수와 마찰력의 변화 특성도 얻었다. 또한 파단면을 전자현미경으로 관찰하여 프레팅 피로기구도 검토해 보았다. 이와 같은 결과는 실제 운전온도에서 프레팅을 받는 증기발생기의 구조건전성 평가에 활용 가능할 것으로 생각된다.

• 대한용접접합학회:학술대회논문집
• /
• 대한용접접합학회 1998년도 특별강연 및 추계학술발표 개요집
• /
• 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.

• Nuclear Engineering and Technology
• /
• 제51권7호
• /
• pp.1805-1815
• /
• 2019

While nickel-based alloys have been widely used for power plants due to corrosion resistance and good mechanical properties, during the last couple of decades, failures of nuclear components increased gradually. One of main degradation mechanisms was primary water stress corrosion cracking at dissimilar metal welds of piping and reactor head penetrations. In this context, precise estimation of welding effects became an important issue for ensuring reliability of them. The present study deals with a series of finite element analyses and crack growth rate evaluation of Alloys 690/152. Firstly, variation of residual stresses and equivalent plastic strains was simulated taking into account welding of a cylindrical block. Subsequently, extraction and pre-cracking of compact tension (CT) specimens were considered from different locations of the block. Finally, crack growth curves of the alloys and heat affected zone were developed based on analyses results combined with experimental data in references. Characteristics of crack growth behaviors were also discussed in relation to mechanical and fracture parameters.

• 한국주조공학회지
• /
• 제13권6호
• /
• pp.532-539
• /
• 1993

The effect of casting variables for making pure Al and Al-5wt%Si alloy rods free from any surface and inside defects was studied by adopting the horizontal continuous casting method with heated mold. The rods were cast under the casting conditions such as the mold temperature of $670{\sim}690^{\circ}C$, water flow rate of $0.2{\sim}0.6{\ell}/min$, and rod diameter of $4{\sim}8mm$, when the melt temperature and mold to cooler distance was fixed at $700^{\circ}C$ and 20mm, respectively. The results represented that the casting speed for good quality rod increased as the water flow rate increased, whereas, the casting speed decreased as the rod diameter or mold temperature increased. The statistical analysis of $2^3$ factorial design was also applied and the results represented that the averaged optimum casting speed for pure Al(302mm/min) was higher than that of Al-5wt%Si alloy(273mm/min) resulting from the difference of superheat applied. The effect of rod diameter on the optimum casting speed was the highest for pure Al as well as Al-5wt%Si alloy. The effect of water flow rate and mold temperature on the optimum casting speed was in decreasing order.

• Nuclear Engineering and Technology
• /
• 제45권7호
• /
• pp.929-940
• /
• 2013

Environmental fatigue of the metallic components in light water reactors has been the subject of extensive research and regulatory interest in Korea and abroad. Especially, it was one of the key domestic issues for the license renewal of operating reactors and licensing of advanced reactors during the early 2000s. To deal with the environmental fatigue issue domestically, a systematic test program has been initiated and is still underway. The materials tested were SA508 Gr.1a low alloy steels, 316LN stainless steels, cast stainless steels, and an Alloy 690 and 52M weld. Through tests and subsequent analysis, the mechanisms of reduced low cycle fatigue life have been investigated for those alloys. In addition, the effects of temperature, dissolved oxygen level, and dissolved hydrogen level on low cycle fatigue behaviors have been investigated. In this paper, the test results and key analysis results are briefly summarized. Finally, an on-going test program for hot-bending of 347 stainless steel is introduced.

• Nuclear Engineering and Technology
• /
• 제47권7호
• /
• pp.895-906
• /
• 2015

Background: Mitigation of primary water stress corrosion cracking (PWSCC) is a significant issue in the nuclear industry. Advanced nickel-based alloys with lower susceptibility have been adopted, although they do not seem to be entirely immune from PWSCC during normal operation. With regard to structural integrity assessments of the relevant components, an accurate evaluation of crack growth rate (CGR) is important. Methods: For the present study, the extended finite element method was adopted from among diverse meshless methods because of its advantages in arbitrary crack analysis. A user-subroutine based on the strain rate damage model was developed and incorporated into the crack growth evaluation. Results: The proposed method was verified by using the well-known Alloy 600 material with a reference CGR curve. The analyzed CGR curve of the alternative Alloy 690 material was then newly estimated by applying the proven method over a practical range of stress intensity factors. Conclusion: Reliable CGR curves were obtained without complex environmental facilities or a high degree of experimental effort. The proposed method may be used to assess the PWSCC resistance of nuclear components subjected to high residual stresses such as those resulting from dissimilar metal welding parts.

• 열처리공학회지
• /
• 제10권4호
• /
• pp.219-231
• /
• 1997

Inconel alloys used as nuclear power plant components have experienced intergranular stress corrosion cracking problems inspite of their good corrosion characteristics. In order to investigate the effects of heat treatments on carbide precipitation and intergranular stress corrosion cracking(IGSCC) in Inconel alloys, DSC(Differential Scanning Calorimeter), TEM, EDXS and static potential corrosion tests were carried out. Thermal treatment at $750^{\circ}C$ for 15hours in Inconel alloys increased the density of intergranular carbide. The carbides are mainly $Cr_7C_3$ in Inconel 600, and $Cr_{23}C_6$ in Inconel 690. The Cr depletion around grain boundary is not crucial factor on IGSCC. The carbides in grain boundary play an important role as acting dislocation source, and as decreasing stress around growing crack.

• 열처리공학회지
• /
• 제14권4호
• /
• pp.220-227
• /
• 2001

The static creep behaviors of dispersion strengthened copper GlidCop were investigated over the temperature range of $650{\sim}690^{\circ}C$ (0.7Tm) and the stress range of 40~55 MPa (4.077~5.61 $kg/mm^2$). The stress exponents for the static creep deformation of this alloy was 8.42, 9.01, 9.25, 9.66 at the temperature of 690, 677, 663, and $650^{\circ}C$, respectively. The stress exponent, (n) increased with decreasing the temperature and became dose to 10. The apparent activation energy for the static creep deformation, (Q) was 374.79, 368.06, 361.83, and 357.61 kg/mole for the stress of 40, 45, 50, and 55 MPa, respectively. The activation energy (Q) decreased with increasing the stress and was higher than that of self diffusion of Cu in the dispersion strengthened copper. In results, it can be concluded that the static creep deformation for dispersion strengthened copper was controlled by the dislocation climb over the ranges of the experimental conditions. Larson-Miller parameter (P) for the crept specimens for dispersion strengthened copper under the static creep conditions was obtained as P=(T+460)(logtr+23). The failure plane observed for SEM slightly showed up transgranular at that experimental range, however, universally it was dominated by characteristic of the intergranular fracture.

• Nuclear Engineering and Technology
• /
• 제46권5호
• /
• pp.633-640
• /
• 2014

In this research, the surface roughness affecting the pressure drop in a pipe used as the steam generator of a PWR was studied. Based on the CFD (Computational Fluid Dynamics) technique using a commercial code named ANSYS-FLUENT, a straight pipe was modeled to obtain the Darcy frictional coefficient, changed with a range of various surface roughness ratios as well as Reynolds numbers. The result is validated by the comparison with a Moody chart to set the appropriate size of grids at the wall for the correct consideration of surface roughness. The pressure drop in a full-scale U-shaped pipe is measured with the same code, correlated with the surface roughness ratio. In the next stage, we studied a reduced scale model of a U-shaped heat pipe with experiment and analysis of the investigation into fluid-structure interaction (FSI). The material of the pipe was cut from the real heat pipe of a material named Inconel 690 alloy, now used in steam generators. The accelerations at the fixed stations on the outer surface of the pipe model are measured in the series of time history, and Fourier transformed to the frequency domain. The natural frequency of three leading modes were traced from the FFT data, and compared with the result of a numerical analysis for unsteady, incompressible flow. The corresponding mode shapes and maximum displacement are obtained numerically from the FSI simulation with the coupling of the commercial codes, ANSYS-FLUENT and TRANSIENT_STRUCTURAL. The primary frequencies for the model system consist of three parts: structural vibration, BPF(blade pass frequency) of pump, and fluid-structure interaction.

• 한국재료학회지
• /
• 제4권8호
• /
• pp.847-854
• /
• 1994

Mg-Zn-Ce계 합금에서 비정질 단상 및 hcp-Mg입자분산형 비정질합금이 20-40%, Zn, 0-10%Ce과 5-20%Zn, 0-5%Ce 의 조성범위에서 각각 생성되었다. 초미세 hcp-Mg입자분산형 $Mg_{85}Zn_{12}Ce_{3}$비정질합금은 급속응고 또는 급속응고리본의 열처리에 의해 Mg입자의 입경을 4-20nm의 범위로 조절할 수 있었으며, 이 범위에서는 밀착굽힘이 가능할 만큼 충분한 인성을 가지고 있었다. 이 합금의 최대인장강도($\sigma_{B}$)와 파단 연신율($\varepsilon_{f}$)은 hcp-Mg입자의 체적분율에 따라서 670-930MPa, 5.2-2.0%의 범위였으며, 최대 비강도($\sigma_{B}$밀도 =$\sigma_{s}$)는 $3.6 \times 10^5N \cdot m/kg$에 달하였다. 이와 같이 Mg입자분산형 비정질 합금의($\sigma_{B}$), ($\sigma_{s}$)그리고 $\varepsilon_{f}$의 최대치가 Mg-Zn-Ce계 비정질합금(690MPa, $2.5 \times 10^5N \cdot m/kg$, 2.5%)보다 월등하게 높다는 것은 주목할 만 하다. 복합상 조직이 형성됨으로서 기계적 강도가 증가하는 것은 동일 조성의 비정질상보다 강한 hcp과포화 고용체의 분산강화에 기인하는 것이라고 고찰되었다.