• Journal of the Korean institute of surface engineering
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• v.52 no.6
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• pp.293-297
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• 2019

Quasi-nano-hardness and corrosion behaviors of neutron-irradiated reactor pressure vessel (RPV) steels such as 15Ch2MFA (Ni<0.4, 2.520 n/㎠ (En>1.0 MeV) for 32 days. Quasi-nano-hardnesses of the 15Ch2MFA and 15Cr2NHFA steels were 183.8 and 179.8 Hv, respectively. Their corrosion rates and corrosion potentials were 2.4×10^-4Acm^-2, -515.9 mV_SHE and 6.8×10^-4 Acm^-2, -523.6 mV_SHE in NACE standard TM0284-96 solution at room temperature, respectively. 15Ch2MFA steel showed better quasi-nano-hardness and corrosion resistance than 15Cr2NHFA steel in this test condition.

• Journal of Ocean Engineering and Technology
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• v.7 no.2
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• pp.91-102
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• 1993

The elastic-plastic fracture thoughness J sub(IC) of Nuclear Reactor Vessel Steel(SA 508-3) which has high toughness was discussed at temperatures RT, $-20^{\circ}C$, $200^{\circ}C$ and 1/2/CT specimen was used for this study. Especially the two methods recommended in ASTM and JSME were compared. It was difficult to find J sub(IC) by ASTM R-curve method with the specimen used for this research, while JSME R-curve method yielded good result. The stretched zone width menthod gave slightly larger J sub(IC) values than those by the R-curve method for SA 508-3 steel.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.413-422
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• 2014

Presented in this study are the variations of microstructures and mechanical properties with tempering and Post-Weld Heat Treatment (PWHT) conditions for SA508 Gr.4N steel used as Reactor Pressure Vessel (RPV) material. The blocks of model alloy were austenitized at the conventional temperature of $880^{\circ}C$ then tempered and post-weld heat treated at four different conditions. The hardness and yield strength decrease with increased tempering and PWHT temperatures, but impact toughness is significantly improved, especially in the specimens tempered at $630^{\circ}C$. The sample tempered at $630^{\circ}C$ with PWHT at $610^{\circ}C$ shows optimum mechanical properties in hardness, strength, and toughness, excluding only the transition property in the low temperature region. The microstructural observation and quantitative analysis of carbide size distribution show that the variations of mechanical properties are caused by the under-tempering and carbide coarsening which occurred during the heat treatment process. The introduction of PWHT results in the deterioration of the ductile-brittle transition property by an increase of coarse carbides controlling cleavage initiation, especially in the tempered state at $630^{\circ}C$.

• Structural Engineering and Mechanics
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• v.60 no.2
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• pp.337-350
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• 2016

Even though extensive researches have been performed for steam explosion due to their complex mechanisms and inherent uncertainties, establishment of severe accident management guidelines and strategies is one of state-of-the arts in nuclear industry. The goal of this research is primarily to examine effects of vessel failure modes and locations on nuclear facilities under typical steam explosion conditions. Both discrete and integrated models were employed from the viewpoint of structural integrity assessment of steel components and evaluation of the cracking and crushing in reinforced concrete structures. Thereafter, comparison of systematic analysis results was performed; despite the vessel failure modes were dominant, resulting maximum stresses at the all steel components were sufficiently lower than the corresponding yield strengths. Two failure criteria for the reinforced concrete structures such as the limiting failure ratio of concrete and the limiting strains for rebar and liner plate were satisfied under steam explosion conditions. Moreover, stresses of steel components and reinforced concrete structures were reduced with maximum difference of 12% when the integrated model was adopted comparing to those of discrete models.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1732-1741
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• 2020

The structural integrity of welded joints in the reactor pressure vessel (RPV) is directly related to the safety of nuclear power plants. The RPV is made from SA508-III steel in a pressurized water reactor. In this study, we investigated the effects of temperature on the tensile and fracture toughness properties of Chinese SA508-III welded joint in different sampling areas in order to provide reference data for structural integrity assessments of RPVs. The specimens used in tensile and fracture toughness tests were fabricated from the base metal (BM), weld metal (WM), and the heat-affected zone (HAZ) in the welded joint. The representative testing temperatures included the ambient temperature (20 ℃), upper shelf temperature (100 ℃), and service temperature (320 ℃). The results showed that temperature greatly affected the fracture toughness (JIC) values for the SA508-III welded joint. The JIC values for BM and HAZ both decreased remarkably from 20 ℃ to 320 ℃. The fracture morphologies showed that the BM and HAZ in the welded joint exhibited fully ductile fracture at 20 ℃, whereas partial cleavage fracture was mixed in ductile fracture mode at 100 ℃ and 320 ℃. The WM exhibited the ductile and cleavage fracture mixed mode at various temperatures, and the JIC values showed slight changes.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.11 no.4
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• pp.160-167
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• 2012

This paper is concerned with a particular fracture mechanics parameter ${\Delta}K_{th}$, known as the 'threshold stress intensity range', or 'fatigue threshold'. This threshold ${\Delta}K_{th}$ constitutes, as it were, a hinge between the notion of crack initiation and the notion of crack growth. It has often been thought that, like the endurance limit, it could be an intrinsic criterion of the material. The study was conducted on a SA-508 pressure vessel steel used in the nuclear power industry. This material exhibits a typical threshold effect in the range of the crack growth rates which were determined; that is, below approximately $da/dN=10^{-6}mm/cycle$, the slope of the da./dN versus ${\Delta}K$ curve is almost vertical. The value of ${\Delta}K_{th}$ was determined at a growth rate of $10^{-7}$ mm/cycle according to the ASTM Standard for threshold testing. The fatigue threshold values are in the range 21 $kg/mm^{3/2}$ to 12 $kg/mm^{3/2}$ depending on the stress ratio effect.

• Korean Journal of Metals and Materials
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• v.48 no.3
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• pp.194-202
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• 2010

Application of a stronger and more durable material for reactor pressure vessels (RPVs) might be an effective way to insure the integrity and increase the efficiency of nuclear power plants. A series of research projects to apply the SA508 Gr.4 steel in ASME code to RPVs are in progress because of its excellent strength and durability compared to commercial RPV steel (SA508 Gr.3 steel). In this study, the microstructural characteristics and mechanical properties of SA508 Gr.3 Mn-Mo-Ni low alloy steel and SA508 Gr.4N Ni-Mo-Cr low alloy steel were investigated. The differences in the stable phases between these two low alloy steels were evaluated by means of a thermodynamic calculation using ThermoCalc. They were then compared to microstructural features and correlated with mechanical properties. Mn-Mo-Ni low alloy steel shows the upper bainite structure that has coarse cementite in the lath boundaries. However, Ni-Mo-Cr low alloy steel shows the mixture of lower bainite and tempered martensite structure that homogeneously precipitates the small carbides such as $M_{23}C_6$ and $M_7C_3$ due to an increase of hardenability and Cr addition. In the mechanical properties, Ni-Mo-Cr low alloy steel has higher strength and toughness than Mn-Mo-Ni low alloy steel. Ni and Cr additions increase the strength by solid solution hardening. In addition, microstructural changes from upper bainite to tempered martensite improve the strength of the low alloy steel by grain refining effect, and the changes in the precipitation behavior by Cr addition improve the ductile-brittle transition behavior along with a toughening effect of Ni addition.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.74-83
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• 1995

The accurate determination of the fast neutron flux/fluence onto the pressure vessel is an essential part of the reactor pressure vessel surveillance program. It has been reported recently that the iron cross section data in ENDF/B versions III through V might underestimate the flux/fluence of fast neutrons in steel structures such as reactor pressure vessel. In this study, for the comparison of iron data of ENDF/B-IV and VI we produced two 47-group cross section sets, CXFe-IV and CXFe-Ⅵ, which are based on Yonggwang nuclear unit-3/4 model and the iron data of ENDF/B-IV and VI, respectively. A comparison was made of the results obtained from DOT4.3 calculation using CXFe-IV and CXFe-VI. From the results, it was found that the fast flux(E 〉 1.0 MeV), which is important for the pressure vessel embrittlement analysis, increases by about 7.6％ at the inner wall and 20％ at the outer wall of the vessel, if the iron data are used from ENDF/B-VI instead of ENDF/B-IV.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.182-187
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• 1996

Deformation characteristics of miniature plate tensile specimens have been studied to develop the thickness requirement and a correlation to estimate the mechanical properties of bulk material from miniature specimen data. The material used was a SA 508 C1.3 reactor pressure vessel steel and the thicknesses of miniature tensile specimens varied from ().12 m to 2 mm. The effects of thickness on the tensile deformation properties such as strength, ductility, and necking characteristics were analyzed. The yield and ultimate tensile strengths were independent of specimen thickness when the thickness was larger than about 0.2 mm. The uniform and total elongations decreased as the specimen thickness decreased. It was also observed that the uniform strain component in the width direction decreased with decrease in the specimen thickness, however, that in the thickness direction was rather constant in total thickness range studied. Based on this observation and a relationship between the necking angle and the ratio between strain components, a correlation between the uniform elongations of miniature specimen and standard specimen was derived. The uniform elongations calculated by this new correlation agreed well with the measured values.

• Korean Journal of Metals and Materials
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• v.46 no.12
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• pp.771-779
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• 2008

An effective way of increasing the strength and fracture toughness of reactor pressure vessel steels is to change the material specification from that of Mn-Mo-Ni low alloy steel(SA508 Gr.3) to Ni-Mo-Cr low alloy steel(SA508 Gr.4N). In this study, we evaluate the effects of alloying elements on the microstructural characteristics of Ni-Mo-Cr low alloy steel. The changes in the stable phase of the SA508 Gr.4N low alloy steel with alloying elements were evaluated by means of a thermodynamic calculation conducted with the software ThermoCalc. The changes were then compared with the observed microstructural results. The calculation of Ni-Mo-Cr low alloy steels confirms that the ferrite formation temperature decreases as the Ni content increases because of the austenite stabilization effect. Consequently, in the microscopic observation, the lath martensitic structure becomes finer as the Ni content increases. However, Ni does not affect the carbide phases such as $M_{23}C_6 $ and $M_7C_3$. When the Cr content decreases, the carbide phases become unstable and carbide coarsening can be observed. With an increase in the Mo content, the $M_2C$ phase becomes stable instead of the $M_7C_3$ phase. This behavior is also observed in TEM. From the calculation results and the observation results of the microstructure, the thermodynamic calculation can be used to predict the precipitation behavior.