• 방사성폐기물학회지
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• 제14권2호
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• pp.91-100
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• 2016

경수로 원전을 대상으로 원전 내 방사화 대상 물질인 스테인리스강, 탄소강 및 콘크리트의 불순물 정보 적용여부에 따른 방사화 핵종 재고량을 계산하였다. 본 연구에서 탄소강은 압력용기 물질에 사용되었고, 스테인리스강은 압력용기 내부 물질에 사용되었으며, 일반 콘크리트가 생체 차폐체에 사용되었다. 금속 물질에 대해서는 참고자료 1개의 불순물 함량 정보를 적용하였고, 콘크리트 물질에서는 참고자료 5개의 불순물 함량 정보를 적용하여 평가를 수행하였다. 방사화 핵종 재고량 전산해석 시 중성자속 계산에는 MCNP 전산코드를, 방사화 계산에는 FISPACT 전산코드를 각각 사용하였다. 계산 결과, 금속 물질에서 불순물을 포함한 경우가 그렇지 않은 경우보다 비방사능이 2배 이상 높았으며, 특히 콘크리트에서는 불순물을 포함한 경우가 그렇지 않은 경우보다 최대 30배 이상 비방사능이 높게 계산되었다. 방사화 핵종의 생성반응과 재고량을 분석한 결과, 금속 구조물에서는 불순물 중 Co원소와 중성자에 의해 생성되는 방사화 핵종인 Co-60이, 콘크리트에서는 불순물 중 Co, Eu 원소와 중성자에 의해 생성되는 방사화 핵종인Co-60, Eu-152, Eu-154 이 방사성폐기물 준위 결정에 큰 영향을 미치고 있음을 확인하였다. 본 연구의 결과는 원전 해체 계획 수립 시 방사화 핵종 재고량 평가 및 규제에 활용될 수 있을 뿐 아니라, 해체를 고려한 원전 또는 원자력시설의 설계 단계에서도 참고자료로 활용 될 것으로 판단된다.

• 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.

• 한국해양공학회지
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• 제9권1호
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• pp.111-119
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• 1995

The elastic-plastic fracture toughness($J_{IC}$) and fracture resistance (J-R curve) of SA508-3 alloy steel used for nuclear reactor pressure vessel are investigated by using CT-type specimens. Fracture toughness tests are conducted by unloading compliance method and multiple specimen method at room temperature, -2$0^{\circ}C$ and 20$0^{\circ}C$. The apparent negative crack growth phenomenon which usually arises in partial unloading compliance test is well known. The negative crack growth phenomenon in determining J sub(IC) or J-R cure from partial unloading compliance experiments may be eliminated by the offset technique. In this study, the evaluation of $J_{IC}$ multiple specimen method recommended by the JSME gives the most reliable results by using half-size CT(similar-type) specimens.

• 한국해양공학회지
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• 제14권1호
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• pp.44-51
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• 2000

Safety and integrity are required for reactor pressure vessels because they are operated in high temperature. There are single specimen method multiple specimen method and load ratio analysis method which used as evaluation of safety and integrity for reactor pressure vessels. In this study the fracture resistance curve(J-R curve) elastic-plastic fracture toughness($J_{IC}$) and material tearing modulus ($T_{mat}$) of SA 508 class 3 alloy steel used as reactor pressure vessel steel are measured and evaluated at room temperature 20$0^{\circ}C$ and 30$0^{\circ}C$ according to unloading compliance method and load ration analysis method. And then the comparison with experimental $J_{IC}$ and theoretical$J_{IC}$ by local fracture strain is managed.

• 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.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.357-373
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• 2022

The high-temperature gas-cooled reactor pebble-bed module project is the first commercial Generation-IV NPP(Nuclear Power Plant) in China. A new joint is used for the vertical support of RPV(Reactor Pressure Vessel). The steel corbel is integrally embedded into the reactor-cabin wall through eight asymmetrically arranged pre-stressed high-strength bolts, achieving the different path transmission of shear force and moment. The vertical monotonic loading test of two specimens is conducted. The results show that the failure mode of the joint is bolt fracture. There is no prominent yield stage in the whole loading process. The stress of bolts is linearly distributed along the height of corbel at initial loading. As the load increases, the height of neutral axis of bolts gradually decreases. The upper and lower edges of the wall opening contact the corbel plate to restrict the rotation of the corbel. During the loading, the pre-stress of some bolts decreases. The increase of the pre-stress strength ratio of bolts has no noticeable effect on the structure stiffness, but it reduces the ultimate bearing capacity of the joint. A simplified calculation model for the elastic stage of the joint is established, and the estimation results are in good agreement with the experimental results.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.1109-1112
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• 2017

The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 1998년도 특별강연 및 춘계학술발표 개요집
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• pp.140-143
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• 1998

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.920-926
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• 2021

Capsules containing NaCl-MgCl₂ salt with 316L stainless steel or alloy N samples were irradiated in the Ohio State University Research Reactor for 21 nonconsecutive hours. A custom irradiation vessel was designed for this purpose, and details on its design and construction are given. Stainless steel samples that were irradiated during exposure had less corrosive attack than samples exposed to the same conditions without irradiation. Alloy N samples showed no significant effect of irradiation. This work shows a method for conducting in-reactor irradiation-corrosion experiments in static molten salts and presents preliminary data showing that neutron irradiation may decelerate corrosion of alloys in molten chloride salts.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.225-225
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• 1999