• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.543-550
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• 2012

The decrease in the fracture toughness of ferritic steels in a reactor pressure vessel is an important factor in determining the lifetime of a nuclear power plant. A surveillance program has been in place in Korea since 1979 to assess the structural integrity of RPV steels. In this work, the surveillance data were collected and analyzed statistically in order to derive the empirical relationship between the embrittlement and strengthening of irradiated reactor pressure vessel steels. There was a linear relationship between the yield strength change and the transition temperature shift change at 41 J due to irradiation. The proportional coefficient was about $0.5^{\circ}C$/MPa in the base metals (plate/forgings). The upper shelf energy decrease ratio was non-linearly proportional to the yield strength change, and most of the data lay along the trend curve of the US results. The transition regime temperature interval, ${\Delta}T_T$, was less than the US data. The overall change from irradiation was very similar to the US results. It is expected that the results of this study will be applied to basic research on the multiscale modeling of the irradiation embrittlement of RPV materials in Korea.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.290.2-290
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• 2002

• 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.

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.203-213
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• 1988

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.294-294
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.292.1-292
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.367.2-367.2
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• 2003

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.100-106
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• 2014

There is a growing need to introduce advanced pressure vessel steels with higher strength and toughness for the optimizatiooCn of the design and construction of longer life and larger capacity nuclear power plants. SA508 Gr.4N Ni-Cr-Mo low alloy steels have superior strength and fracture toughness, compared to SA508 Gr.3 Mn-Mo-Ni low alloy steel. Therefore, the application of SA508 Gr.4N low alloy steel could be considered to satisfy the strength and toughness required in advanced nuclear power plants. The purpose of this study is to characterize the microstructure and mechanical properties of SA508 Gr.4N low alloy steels. 1 ton ingot of SA508 Gr.4N model alloy was fabricated by vacuum induction melting followed by forging, quenching, and tempering. The predominant microstructure of the SA508 Gr.4N model alloy is tempered martensite having small packet and fine Cr-rich carbides. The yield strength at room temperature was 540MPa, and it was decreased with an increase of test temperature while DSA phenomenon occurred at around $288^{\circ}C$. Overall transition property of SA508 Gr.4N model alloy was much better than SA508 Gr.3 low alloy steel. The index temperature, $T_{41J}$, of SA508 Gr.4N model alloy was $-132^{\circ}C$ in Charpy impact tests, and reference nil-ductility transition temperature, $RT_{NDT}$ of $-105^{\circ}C$ was obtained from drop weight tests. From the fracture toughness tests performed in accordance with the ASTM standard E1921 Master curve method, the reference temperature, $T_0$ was $-147^{\circ}C$, which was improved more than $60^{\circ}C$ compared to SA508 Gr.3 low alloy steels.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.113-121
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• 2014

In this work, an integrated model including molecular dynamics and chemical rate theory was implemented to calculate the growth of point defect clusters(PDC) and copper-rich precipitates(CRP) which could change the mechanical properties of reactor pressure vessel(RPV) steels in a nuclear power plant. A number of time-dependent differential equations were established and numerically integrated to estimate the evolution of irradiation defects. The calculation showed that the concentration of the vacancies was higher than that of the self-interstitial atoms. The higher concentration of vacancies induced a formation of the CRPs in the later stage. The size of the CRPs was used to estimate the mechanical property changes in RPV steels, as is the same case with the PDCs. The calculation results were compared with the measured values of yield strength change and Charpy V-notch transition temperature shift, which were obtained from the surveillance test data of Korean light water reactors(LWRs). The estimated values were in fair agreement with the experimental results in spite of the uncertainty of the modeling parameters.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.129-136
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• 1997

The automated ball indentation(ABI) test was utilized to develop a semi-nondestructive method for estimating the fracture toughness( $K_{JC}$ ) in the transition temperature range. The key concept of the method is that the indentation deformation energy to the load at which the mean ball-specimen contact pressure reaches the fracture stress is related to the fracture energy of the material. ABI tests were performed for the reactor pressure vessel(RPV) base and weld metals at the temperatures of-15$0^{\circ}C$~$0^{\circ}C$ and the fracture toughness (estimated $K_{JC}$ ) was calculated from the indentation load-depth data. For all steels the temperature dependence of the estimated fracture toughness was almost the same as that ASTM $K_{JC}$ master curve The reference temperatures( $T_{o}$)of the steels were determined form the estimated $K_{JC}$ versus temperature curves. The reference temperature was well correlated with the index temperature of 41J Charpy impact energy( $T_{41J}$).).).