• 한국압력기기공학회 논문집
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• 제10권1호
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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.

• 열처리공학회지
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• 제25권3호
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• pp.121-125
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• 2012

The large forgings used in chemical plants or nuclear power plants are produced by complex heat treatment. because of thickness up to 200~300 mm and weight up to 200~300 ton, setting proper heat treatment cycle is so difficult. In addition, defects of products make companies wasting large money and valuable time. In this study, to reduce try & err, when setting heat treatment of reactor pressure vessel steel SA508Gr.3, carrying out the basic mechanical property test of SA508 Gr.3 and testing hardness of SA508Gr.3 in various tempering temperature. and calculating temper curve with Hollomon-Jaffe parameter.

• 대한기계학회논문집A
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• 제39권4호
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• pp.383-390
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• 2015

본 논문에서는 원전 배관 재료의 변형거동에 미치는 하중속도의 영향을 파악하기 위해서, SA508 Gr.1a 저합금강과 SA312 TP316 스테인리스강 배관재를 대상으로 상온과 원전 운전온도인 $316^{\circ}C$에서 다양한 변형률 속도로 인장시험을 수행하였다. 시험 결과, 상온에서는 두 배관재의 변형거동이 일반적인 하중속도 의존성을 보였다. 즉, 하중속도가 증가함에 따라 강도는 증가하고 연성은 감소하는 경향을 보였다. 그러나, $316^{\circ}C$에서는 하중속도가 증가함에 따라 SA508 Gr.1a 저합금강의 강도와 연신률이 모두 감소하였고 단면수축률은 감소 후 증가하는 비선형 거동을 보였다. SA312 TP316 스테인리스강의 강도, 연신률, 그리고 단면수축률은 하중속도에 관계없이 거의 일정한 값을 보였다. 시험 온도에 관계없이 SA508 Gr.1a 저합금강의 가공경화능력은 하중속도에 거의 영향을 받지 않았으나, SA312 TP316 스테인리스강에서는 하중속도가 증가함에 따라 가공경화능력이 감소하였다.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.773-780
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• 2012

The article presented is concerned with an evaluation of the corrosion behavior of SA-508 low alloy steel (LAS) and Type 309L stainless steel (SS) cladding of a reactor pressure vessel under the simulated primary water chemistry of a pressurized water reactor (PWR). The uniform corrosion and galvanic corrosion rates of SA-508 LAS and Type 309L SS were measured in three different control conditions: power operation, shutdown, and power operation followed by shutdown. In all conditions, the dissimilar metal coupling of SA-508 LAS and Type 309L SS exhibited higher corrosion rates than the SA-508 base metal itself due to severe galvanic corrosion near the cladding interface, while the corrosion of Type 309L in the primary water environment was minimal. The galvanic corrosion rate of the SA-508 LAS and Type 309L SS couple measured under the simulated power operation condition was much lower than that measured in the simulated shutdown condition due to the formation of magnetite on the metal surface in a reducing environment. Based on the experimental results, the corrosion rate of SA-508 LAS clad with Type 309L SS was estimated as a function of operating cycle simulated for a typical PWR.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.42-48
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• 2020

The effect of improving the tensile and J-R fracture toughness properties of SA508 Gr.1a on the LBB margin for the main steam pipe is investigated. The material properties and microstructure images of the existing main steam piping material SA106 Gr.C used in domestic nuclear power plants and the newly selected material SA508 Gr.1a were compared. For each material, LBB margins were calculated and compared through finite element analysis and crack instability evaluation. The LBB margin of the improved SA508 Gr.1a is found to be greatly improved compared to that of the existing SA106 Gr.C and SA508 Gr.1a. This is because of the increased material's strength and J-R fracture toughness compared to the previous materials. In order to analyze the effect of physical property change on the LBB margin, the sensitivity of each LBB margin according to the variation of tensile strength and J-R fracture toughness was analyzed. The effect of the change in tensile strength was found to be greater than that of the change in fracture toughness. Therefore, an increase in strength significantly influenced the improvement of the LBB margin of the improved SA508 Gr.1a.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.330-330
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.203.2-203
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• 2001

• Journal of Welding and Joining
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• 제13권2호
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• pp.139-149
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• 1995

Many pressure vessels for the power plants are fabricated from low alloy ferritic steels. The inner sides of the pressure vessels are commonly weld_cladded with austenitic stainless steels to minimize problems of corrosive attack. The submerged-arc welding(SAW) process is now used in preference to other processes because of the possibilities open to automation to reduce the overaII welding times. The most reliable way to avoid underclad cracks(UCC) which are often detected at the overlap of the clad beads is to use nonsusceptible steels such as SA508 class 3. At present domestically developed forging steel of SA508 cl.S is now being cladded with single layer by using 90mm wide strip, which transfers higher heat input into the base metal compared to the conventional two layers strip cladding which has been in wide use with 30-60 mm wide strip. But the current indices for the influence of heat input on crack susceptibility are not accurate enough to express the subtle difference in crack susceptibility of the steel. Therefore, the purpose of this present study is: l) To determine UCC susceptibility on domestic forging steel, SA508 cl.S cladded with single layer by using submerged arc 90mm strip and, 2) To optimize heat input range by which the crack susceptibility could be eliminated.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2004년도 춘계 학술발표대회 개요집
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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)

• 한국해양공학회지
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• 제7권2호
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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.