• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2610-2615
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• 2021

The prediction of irradiation-induced transition temperature shift for RPV steels is an important method for long term operation of nuclear power plant. Based on the irradiation embrittlement data, an irradiation-induced transition temperature shift prediction model is developed with machine learning method XGBoost. Then the residual, standard deviation and predicted value vs. measured value analysis are conducted to analyze the accuracy of this model. At last, Cu content threshold and saturation values analysis, temperature dependence, Ni/Cu dependence and flux effect are given to verify the reliability. Those results show that the prediction model developed with XGBoost has high accuracy for predicting the irradiation embrittlement trend of RPV steel. The prediction results are consistent with the current understanding of RPV embrittlement mechanism.

• Nuclear Engineering and Technology
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• 제44권5호
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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.

• 한국재료학회지
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• 제10권4호
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• pp.305-311
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• 2000

샤피 V-노치 충격 하중-변위 곡선으로부터 얻은 균열정지하중을 이용하여 원자로압력용기강의 균열정지파괴인성($K_{Ia}$)을 예측할 수 있는 방법을 모색하고 그 타당성을 고찰하였다. 샤피충격 하중-변위 곡선으로부터 얻은 균열정지하중값의 변화는 특성온도로 보정된 지수함수의 형태로 잘 표현될 수 있었다. 특성온도 $T_{Pa=2kN}$은 실험적인 무연성천이온도($T_{NDT}$) 및 $T_{41\;J}$과 높은 상관성을 나타냈으며, 원자로압력용기강의 균열정지파괴인성을 표현하는 새로운 특성온도로 사용할 수 있을 것으로 판단되었다. 또한 균열정지하중값의 변화는 파면으로부터 측정된 안정균열길이의 변화와 매우 높은 상관성을 나타내었다. 따라서 무딘 노치를 갖는 시편에 대한 계장화샤피충격시험을 통하여 균열정지하중 및 안정균열길이를 측정하믈써 비교적 정확하게 원자로압력용기강에 대한 하한값의 파괴인성치($K_{Ia}$)를 평가하는 것이 가능한 것으로 판단되었다.

• 한국표면공학회지
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• 제52권6호
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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.

• 대한금속재료학회지
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• 제47권9호
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• pp.533-541
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• 2009

Materials used for a reactor pressure vessel(RPV) are required high strength and toughness, which determine the safety margin and life of a reactor. Ni-Mo-Cr low alloy steel shows better mechanical properties than existing RPV steels due to higher Ni and Cr contents compared to the existing RPV steels. The present study focuses on effects of Ni, Cr contents on the cleavage fracture toughness of Ni-Mo-Cr low alloy steels in the transition temperature region. The fracture toughness was characterized by a 3-point bend test of precracked Charpy V-notch(PCVN) specimens based on ASTM E1921-08. The test results indicated that the fracture toughness was considerably improved with an increase of Ni and Cr contents. Especially, control of Cr content was more effective in improving fracture toughness than manipulating Ni content, though Charpy impact toughness was changed more extensively by adjusting Ni content. These differences between changes in the fracture toughness and that in the impact toughness were derived from microstructural features, such as martensite lath size and carbide precipitation behavior.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.208-208
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• 1998

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.271.1-271
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• 1999

• 한국압력기기공학회 논문집
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• 제10권1호
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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.

• 대한금속재료학회지
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• 제48권2호
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• pp.122-132
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• 2010

Higher strength and fracture toughness of reactor pressure vessel steels can be obtained by changing 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). However, the operation temperature of the reactor pressure vessel is more than $300^{\circ}C$ and the reactor operates for over 40 years. Therefore, we need to have phase stability in the high temperature range in order to apply the SA508 Gr.4N low alloy steel for a reactor pressure vessel. It is very important to evaluate the temper embrittlement phenomena of SA508 Gr.4N for an RPV application. In this study, we have performed a Charpy impact test and tensile test of SA508 Gr.4N low alloy steel with changing impurity element contents such as Mn and P. And also, the mechanical properties of these low alloy steels after longterm heat treatment ($450^{\circ}C$, 2000hr) are evaluated. Further, evaluation of the temper embrittlement by fracture analysis was carried out. Temper embrittlement occurs in KL4-Ref and KL4-P, which show a decrease of the elongation and a shifting of the transition curve toward high temperature. The reason for the temper embrittlement is the grain boundary segregation of the impurity element P and the alloying element Ni. However, KL4-Ref shows temper embrittlement phenomena despite the same contents of P and Ni compared with SC-KL4. This result may be caused by the Mn contents. In addition, the behavior of embrittlement is not largely affected by the formation of $M_3P$ phosphide or the coarsening of Cr carbides.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.501-509
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• 2019

In this paper a body-centered cubic(BCC) crystal plasticity model based on microscopic dislocation mechanism is introduced and numerically implemented. The model is coupled with irradiation effect via tracking dislocation loop evolution on each slip system. On the basis of the model, uniaxial tensile tests of unirradiated and irradiated RPV steel(take Chinese A508-3 as an example) at different temperatures are simulated, and the simulation results agree well with the experimental results. Furthermore, crystal plasticity damage is introduced into the model. Then the damage behavior before and after irradiation is studied using the model. The results indicate that the model is an effective tool to study the effect of irradiation and temperature on the mechanical properties and damage behavior.