• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.15 no.1
• /
• pp.18-27
• /
• 2019

The Advanced Power Reactor 1400 (APR1400) Steam Generator (SG) uses alloy 690 as a tube material and SA-508 Grade 3 Class 1 as a tubesheet material to form tube-to-tubesheet joint through hydraulic expansion process. In this paper, the residual stresses in the SG tube-to-tubesheet contact area was investigated by applying Model-Based System Engineering (MBSE) methodology and the V-model. The use of MBSE transform system description into diagrams which clearly describe the logical interaction between functions hence minimizes the risk of ambiguity. A theoretical and Finite Element Methodology (FEM) was used to assess and compare the residual stresses in the tube-to-tubesheet contact area. Additionally, the axial strength of the tube to tubesheet joint based on the pull-out force against the contact joint force was evaluated and recommended optimum autofrettage pressure to minimize residual stresses in the transition zone given. A single U-tube hole and tubesheet with ligament thickness was taken as a single cylinder and plane strain condition was assumed. An iterative method was used in FEM simulation to find the limit autofrettage pressure at which pull-out force and contact force are of the same magnitude. The joint contact force was estimated to be 20 times more than the pull-out force and the limit autofrettage pressure was estimated to be 141.85MPa.

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.35 no.8
• /
• pp.951-956
• /
• 2011

Inconel alloys are generally used as steam generator tubes in nuclear power plants. These alloys are highnickel chromium alloys that exhibit excellent resistance to aqueous corrosion. In this paper, the effects of elevated temperatures such as an operating temperature of $320^{\circ}C$ on the fretting fatigue behavior of inconel 600 and 690. We observed that the plain and fretting fatigue limits at $320^{\circ}C$ were slightly lower than those at room temperature. The frictional forces varied depending on the number of load cycles. After each test, we studied the fretting fatigue mechanisms via SEM observations. These results can be used for structural integrity evaluations at elevated temperatures and for studying fretting damage in steam generator systems.

• Nuclear Engineering and Technology
• /
• v.46 no.5
• /
• 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.

• Transactions of Materials Processing
• /
• v.23 no.7
• /
• pp.405-412
• /
• 2014

As a major type of heat exchanger, the steam generator (SG) produces steam from heat energy of a nuclear power plant reactor. The steam produced by the steam generator flows into a turbine, and plays an important role in electric power generation. The heat transfer tubes in the steam generator consist of approximately 10,000 U-shaped tubes, which perform a structural role and act as thermal boundaries. The heat transfer tubes conduct the thermal energy between the primary coolant (about $320^{\circ}C$, $157kgf/cm^2$) obtained from the reactor and the secondary coolant (about $260^{\circ}C$, $60kgf/cm^2$) as part of the secondary system. Recently, the heat transfer tubes in the steam generator of the pressurized water reactor (PWR) are primarily produced from Alloy 600 and Alloy 690 seamless tubes. As a pilot study to find process parameters for the cold U-bending process using rotary draw bending, numerical and experimental investigations were conducted to produce U-shaped tubes from long straight SUS304L seamless tubes. 3D finite element simulations were run using ABAQUS Explicit with consideration of the elastic recovery. The process parameters studied were the angular speed, the operation period and the bending angle. Experimental verifications were conducted to insure the suitability of the final U-shaped configurations with respect to both ovality and wall thickness.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.3 no.1
• /
• pp.31-40
• /
• 2005

High temperature - high pressure apparatus was developed to simulate nickel fewite corrosion products which were main compositions of the radioactive crud in the nuclear power plant. Corrosion product similar to the crud was obtained by a tube accumulator system. Nickel alloy (Inconel 690) and carbon steel (SA106 Gr. C) were corroded at 270 $\^{circ}C$ in the corrosion product generator. Ni ions and Fe ions dissolved by corrosion reaction were able to be transported to the accumulator because the crud generation mechanism was the solubility change with temperature. To evaluate the properties of simulated corrosion products, scanning electron microscope (SEM) observation and EDAX analysis were performed. SEM observation of corrosion product showed the needlelike or crystal structure of oxide depending on precipitating location. The crystal oxide was the nickel ferrite, which was similar to the crud in nuclear power plants.