• Nuclear Engineering and Technology
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• v.49 no.5
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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.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.11-20
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• 2009

Radiation effects on materials are inherently multiscale phenomena in view of the fact that various processes spanning a broad range of time and length scales are involved. A multiscale modeling approach to embrittlement of pressure vessel steels is presented here. The approach includes an investigation of the mechanisms of defect accumulation, microstructure evolution and the corresponding effects on mechanical properties. An understanding of these phenomena is required to predict the behavior of structural materials under irradiation. We used molecular dynamics (MD) simulations at an atomic scale to study the evolution of high-energy displacement cascade reactions. The MD simulations yield quantitative information on primary damage. Using a database of displacement cascades generated by the MD simulations, we can estimate the accumulation of defects over diffusional length and time scales by applying kinetic Monte Carlo simulations. The evolution of the local microstructure under irradiation is responsible for changes in the physical and mechanical properties of materials. Mechanical property changes in irradiated materials are modeled by dislocation dynamics simulations, which simulate a collective motion of dislocations that interact with the defects. In this paper, we present a multi scale modeling methodology that describes reactor pressure vessel embrittlement in a light water reactor environment.

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

• 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.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.311-322
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• 2013

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

• Nuclear Engineering and Technology
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• v.38 no.7
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• pp.619-638
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• 2006

Low temperature irradiation can significantly harden metallic materials and often lead to strain localization and ductility loss in deformation. This paper provides a review on the radiation effects on the deformation of metallic materials, focusing on microscopic and macroscopic strain localization phenomena. The types of microscopic strain localization often observed in irradiated materials are dislocation channeling and deformation twinning, in which dislocation glides are evenly distributed and well confined in the narrow bands, usually a fraction of a micron wide. Dislocation channeling is a common strain localization mechanism observed virtually in all irradiated metallic materials with ductility, while deformation twinning is an alternative localization mechanism occurring only in low stacking fault energy(SFE) materials. In some high stacking fault energy materials where cross slip is easy, curved and widening channels can be formed depending on dose and stress state. Irradiation also prompts macroscopic strain localization (or plastic instability). It is shown that the plastic instability stress and true fracture stress are nearly independent of irradiation dose if there is no radiation-induced phase change or embrittlement. A newly proposed plastic Instability criterion is that the metals after irradiation show necking at yield when the yield stress exceeds the dose-independent plastic instability stress. There is no evident relationship between the microscopic and macroscopic strain localizations; which is explained by the long-range back-stress hardening. It is proposed that the microscopic strain localization is a generalized phenomenon occurring at high stress.

• Journal of Radiation Protection and Research
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• v.32 no.4
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• pp.168-177
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• 2007

A neural network model to evaluate the neutron exposure on the reactor pressure vessel inner diameter was developed. By using the three dimensional synthesis method described in Regulatory Guide 1.190, a simple linear equation to calculate the neutron spectrum on the reactor pressure vessel was constructed. This model can be used in a quick estimation of fast neutron flux which is the most important parameter in the assessment of embrittlement of reactor pressure vessel. This model also used in the selection of an optimum core loading pattern without the neutron transport calculation. The maximum relative error of this model was less than 3.4% compared to the transport calculation for the calculations from cycle 1 to cycle 23 of Kori unit 1.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.118-123
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• 1998

원자로 압력용기 재료로 사용되고 있는 ASTM SA508 및 SA533 계열의 재료는 결정구조가 체심입방격자(bcc)로서 시험온도별 최대흡수에너지(USE)에 대한 선도를 그리면 고온에서는 연성이 크고, 저온에서는 취성이 큰 전형적인 “S”자 형태의 Cv-천이온도곡선으로 나타난다. 그리고 조사전과 조사후의 연성-취성천이온도곡선을 흡수에너지값이 30ft-lb 또는 50ft-lb인 지점에석 비교해보면 재료의 조사취화(radiation embrittlement)현상으로 온도가 높은 쪽으로 이동됨을 알 수 있으며, 이러한 온도의 이동값은 원자로의 운전수명과 밀접한 관련이 있다. 따라서 조사전후의 흡수에 너지값에 따른 온도변화량를 정확하게 산출하기 위해서는 시편의 온도를 조절하는 장치 및 시편을 아주 짧은 시간내에 충격시험기의 앤빌까지 장전하는 장치 둥의 충격시험시스템 구축은 매우 중요하다. 이에 조사계시험시설(IMEF)에서는 원자로 압력용기 감시시험에 대한 충격시험시스템을 구축하였고, 이의 내용은 감시시험 수행에 기준이 되는 ASTM El85-82 및 과학기술처 고시 제 92-20호의 세부내용을 충분하게 만족시키는 것으로 확인되었으며, 이렇게 확인된 내용들은 현재 국내에서 수행되고 있는 감시시험에 적극적으로 활용되고 있다.

• Structural Engineering and Mechanics
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• v.53 no.6
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• pp.1271-1286
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• 2015

Recently, US NRC revised fracture toughness requirements as 10CFR50.61a to reduce the conservatism of 10CFR50.61. However, unlike previous studies relating to the initial regulation, structural integrity evaluations based on the alternative regulation are not sufficient. In the present study, PTS and P-T limit curve evaluations were carried out by using both regulations and resulting data were compared. With regard to the PTS evaluation, the results obtained from the alternative requirements were satisfied with the criterion whereas those obtained from the initial requirements did not meet the criterion. Also, with regard to the P-T limit curve evaluation, operating margin by 10CFR50.61a was greater than that by 10CFR50.61.

• 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.