• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.10 s.253
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• pp.1173-1178
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• 2006

In a primary reactor cooling system(RCS), a dissimilar weld zone exists between cast stainless steel(CF8M) in a pipe and low-alloy steel(SA508 cl.3) in a nozzle. Thermal aging is observed in CF8M as the RCS is exposed for a long period of time to a reactor operating temperature between 290 and $330^{\circ}C$, while no effect is observed in SA508 cl.3. The specimens are prepared by an artificially accelerated aging technique maintained for 300, 1800 and 3600 hrs at $430^{\circ}C$, respectively. The specimens for elastic-plastic fracture toughness tests are according to the process in the thermal notch is created in the heat affected zone(HAZ) of CF8M and deposited zone. From the experiments, the $J_{IC}$ value notched in HAZ of CF8M presented a rapid decrease up to 300 hours at $430^{\circ}C$ and slowly decreased according to the process in the thermal aging time. Also, the $J_{IC}$ value presented a lower value than that of the CF8M base metal. And, the $J_{IC}$ of the deposited zone presented the lowest value of all other cases.

• Proceedings of the KWS Conference
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• 2004.05a
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• pp.141-143
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• 2004

본 실험에서 SA508 CL.3 강재의 서브머지드아크 용접부에 대한 연구로 입열량의 차이에 따라서 인성과 미세조직과의 관계를 조사하였다. 강도가 크면 인성이 작아지고 연성이 작아지는 것을 확인하고, 입열량이 3-5kJ/$\textrm{mm}^2$를 표준으로 하는데, 본 실험에서는 1.6kJ/$\textrm{mm}^2$, 3.2kJ/$\textrm{mm}^2$, 5.0kJ/$\textrm{mm}^2$ 세 가지 조건으로 실험해 보았다. (중략)

• Journal of Welding and Joining
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• v.16 no.6
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• pp.86-96
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• 1998

This study proposes the analysis method of time series ultrasonic signal using the chaotic feature extraction for degradation extent evaluation. Features extracted from time series data using the chaotic time series signal analyze quantitatively degradation extent. For this purpose, analysis objective in this study is fractal dimension, lyapunov exponent, strange attractor on hyperspace. The lyapunov exponent is a measure of the rate at which nearby trajectories in phase space diverge. Chaotic trajectories have at least one positive lyapunov exponent. The fractal dimension appears as a metric space such as the phase space trajectory of a dynamical system. In experiment, fractal correlation) dimensions, lyapunov exponents, energy variation showed values of 2.217∼2.411, 0.097∼ 0.146, 1.601∼1.476 voltage according to degardation extent. The proposed chaotic feature extraction in this study can enhances precision ate of degradation extent evaluation from degradation extent results of the degraded materials (SA508 CL.3)

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.405-410
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• 2006

Microstructural features and their related mechanical property changes in the 309L cladding and the heat affected zone (HAZ) of SA508 cl.3 steel were investigated through the use of TEM, tensile and small punch (SP) tests. The specimens were irradiated at 563 K up to the neutron fluences of $5.79{\times}10^{19}n/cm^2$ (>1MeV). The microstructure of the clad was mainly composed of a fcc ${\gamma}-phase$, a low percentage of bcc ${\delta}-ferrite$, and a brittle ${\sigma}-phase$. Along the weld fusion line there formed a heavy carbide precipitation with a width of $20{\sim}40{\mu}m$, showing preferential cracking during plastic deformation. The yield stress and ductile-to-brittle transition temperature (DBTT) of the irradiated clads increased. The origin of the hardening and the shift of the DBTT are discussed in terms of the irradiation-produced defect clusters of a fine size and brittle ${\sigma}-phase$.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.12
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• pp.1968-1973
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• 2004

Structural degradations are often experienced on the components of nuclear power plants in reactor pressure vessels (RPV) and steam generators (SG) when these components are exposed to high temperature and high pressure for a long period of time. Such conditions result in the change of microstructures and of mechanical properties of materials, which requires an evaluation of the safeguards related to structural integrity. In a primary reactor cooling system (RCS), a dissimilar weld zone exists between cast stainless steel (CF8M) in a pipe and low-alloy steel (SA508 cl.3) in a nozzle. Thermal aging is observed in CF8M as the RCS is exposed for a long period of time under the operating temperature between 290 and 33$0^{\circ}C$. Under the same conditions, it is well known that degradation is not observed in low alloy steel. An investigation of the effect of thermal aging on the various mechanical properties of the dissimilar weld zone is required. The purpose of the present investigation is to find the effect of thermal aging on the dissimilar weld zone. The specimens are prepared by an artificially accelerated aging technique maintained for various times at 43$0^{\circ}C$, respectively. Then, The various mechanical test for the dissimilar welds are performed.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.825-832
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• 1995

The effect of loading variables and temperature on fatigue crack growth rate in SA508 Cl.3 nuclear pressure vessel steel was investigated in air environment Crack growth rate tests on compact tension specimen of thickness 12mm were conducted by using sinusoidal waveform. The crack length was monitored by compliance method. Test conditions were at 0.1 and 0.5 of load ratio, at 1 and 10 Hz of loading frequency, and at room temperature to 40$0^{\circ}C$. At the lower temperatures, the fatigue crack propagation was not affected by the frequency and temperature, while at the higher temperatures above 12$0^{\circ}C$, fatigue crack growth rate increased with decreasing loading frequency and increasing temperature. This accelerated fatigue crack propagation was associated with the increase of oxidation rate at the ahead of crack tip. Fatigue crack growth rate increased with in-creasing the load ratio. The effect of load ratio was more significant at the lower temperature, while the dependence on load ratio decreased with increasing temperature. The sensitivity of load ratio to temperature can be explained by crack closure with the oxidation process.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.867-868
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• 2004

The maximum width of HAZ of SA508치.3 steel produced by overlay RPV cladding was approximately 10 mm and it was composed of variety of microstructures with various grain size and precipitates. In addition, along the weld fusion line there formed a heavy carbide precipitation zone in the width of $20{\sim}30\;{\mu}m$. 2. As the specimen sampling position approached to the weld fusion line, the increase in yield and tensile strength was approximately 90 and 40 MPa, respectively. Meanwhile, the plastic fracture strain reduced from 14 to 8 percent. 3. The lowest SP energy and the highest ductile to brittle transition temperature in the HAZ were observed at the coarse- and fine-grained HAZ.

• Journal of Welding and Joining
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• v.20 no.2
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• pp.82-89
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• 2002

In order to clarify hot cracking phenomena occurred in Inconel 690 welds and it's prevention, in this study, the cracking behavior and the influence of welding variables on cracking in Inconel 690 overlay welds on Mn-Ni-Cr-Mo steel(SA 508 cl.3) for pressure vessel were investigated by using mock-up test. The main results are as follows: The cracks in Inconel 690 overlay welds were mainly generated near the start and the end part of welding beads adjacent to STS 309L welded outside of Inconel 690 welds. Most of the cracks showed typical solidification crack, and also it was assumed that there was possibility of liquation cracking in HAZ. The existence of Nb constituents or concentration of Nb was recognized on the fracture facets of the solidification cracks in the welds by SMAW. Therefore Nb was considered to be the main factor of the solidification cracking. As the weld heat input was more increased and the weld bead length was longer, the extent of cracking was more increased. Moreover the extent of cracking was considerably decreased by changing of welding sequence to the start and the end part of welds. Hot cracking in welds by GTAW was considerably decreased as compared with that of SMAW. And cracks were well generated in the Inconel 690 overlay welds adjacent to 575 309L welds. This means that the hot cracking susceptibility of Inconel 690 welds was largely varied by chemical components and/or compositions of filter metals, base metals and neighboring welds.

• Journal of the Korean Society of Safety
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• v.14 no.1
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• pp.25-32
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• 1999

The objective of this paper is to evaluate the fracture resistance characteristics of SA508 CL.1a carbon steel, TP347 stainless steel and their associated welds manufactured for primary coolant system of Ulchin 3,4 nuclear power plants. The effect of various parameters such as pipe size, welding method, chemical composition, crack plane orientation, metallography and fractography on the material properties were discussed. Test results showed that the effect of pipe size on fracture toughness is negligible while the effect of welding method on fracture toughness is significant. In addition, the drop of fracture toughness in the field fabrication weld of TP347 stainless steel is probably due to the large amount of $\sigma$-phase precipitated on the $\delta$-ferrite boundary and the large size dimples.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.75-83
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• 2017

This study investigated deformation behavior of major nuclear structural materials under cyclic loading conditions via cyclic stress-strain test. The cyclic stress-strain tests were conducted on SA312 TP316 stainless steel and SA508 Gr.3 Cl.1 low-alloy steel, which are used as materials for primary piping and reactor pressure vessel nozzle respectively, under cyclic load with constant strain amplitude and constant load amplitude at room temperature (RT) and $316^{\circ}C$. From the results of tests, the cyclic hardening and softening behavior, stabilized cyclic stress-strain behavior, and ratcheting behavior of both materials were investigated at both RT and $316^{\circ}C$. In addition, appropriate considerations for cyclic deformation behavior in the structural integrity evaluation of major nuclear components under excessive seismic condition were discussed.