• 방사성폐기물학회지
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• 제15권2호
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• pp.181-189
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• 2017

방사성금속폐기물의 관리 옵션들을 안전성, 경제성, 작업자 피폭, 부피 감용 등의 선별 기준을 적용하여 비교 평가하였다. 원전 해체로부터 발생하는 금속폐기물의 관리 옵션에는 무구속 방출, 제한적 재사용, 그리고 직접 처분이 있다. 고려된 각각의 옵션들은 금속폐기물의 절단과 용융에 의한 부피감용을 수반한다. AHP기법을 적용하여 각 옵션들의 순위를 부여하였다. 방사성금속폐기물을 용융하여 금속 잉곳을 제조한 후 제한적 재이용 또는 무구속 방출하는 방안이 가장 효율적인 옵션으로 도출되었다.

• 방사성폐기물학회지
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• 제18권2호
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• pp.113-118
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• 2020

An options study was performed for the treatment of residual elemental sodium in driver plenums following the chopping operation during the pyroprocessing of used nuclear fuel. Given the pending availability of a multi-function furnace for distillation and consolidation operations in the Fuel Conditioning Facility, the furnace was considered for the processing of driver plenums. Although two options (oxidation and distillation) could be performed in the multi-function furnace, neither option has been developed sufficiently to date to warrant the use of the furnace for treatment operations. Thus, it was decided to defer the treatment of elemental sodium from driver plenums in the multi-function furnace until more developed technologies and/or furnaces become available. In the interim, storage of the plenums and characterization efforts are recommended.

• Nuclear Engineering and Technology
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• 제38권6호
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• pp.505-534
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• 2006

Ensuring sufficient supplies of clean, economic and acceptable energy is a critical global challenge for the 21st century. There seems little alternative to a greatly expanded role for nuclear power, but implementation of this option will depend on ensuring that all resulting wastes can be disposed of safely. Although there is a consensus on the fundamental feasibility of such disposal by experts in the field, concepts have to be developed to make them more practical to implement and, in particular, more acceptable to key stakeholders. By considering global trends and using illustrative examples from Japan, key areas for future R&D are identified and potential areas where the synergies of international collaboration would be beneficial are highlighted.

• Nuclear Engineering and Technology
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• 제41권7호
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• pp.953-968
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• 2009

It is important to accurately predict the temperature and density distributions in large stratified enclosures both for design optimization and accident analysis. Current reactor system analysis codes only provide lumped-volume based models that can give very approximate results. Previous scaling analysis has shown that stratified mixing processes in large stably stratified enclosures can be described using one-dimensional differential equations, with the vertical transport by jets modeled using integral techniques. This allows very large reductions in computational effort compared to three-dimensional CFD simulation. The BMIX++ (Berkeley mechanistic MIXing code in C++) code was developed to implement such ideas. This paper summarizes major models for the BMIX++ code, presents the two-plume mixing experiment simulation as one validation example, and describes the codes' application to the liquid salt buffer pool system in the AHTR (Advanced High Temperature Reactor) design. Three design options have been simulated and they exhibit significantly different stratification patterns. One of design options shows the mildest thermal stratification and is identified as the best design option. This application shows that the BMIX++ code has capability to provide the reactor designers with insights to understand complex mixing behavior with mechanistic methods. Similar analysis is possible for liquid-metal cooled reactors.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2565-2571
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• 2020

Background: Understanding the behaviour of nuclear fuel claddings by conducting burst test on single cladding tube under simulated loss-of-coolant accident conditions and developing theoretical cum empirical predictive computer codes have been the focus of several investigations. The developed burst criterion (a) assumes symmetrical deformation of cladding tube in contrast to experimental observation (b) interpolates the properties of Zircaloy-4 cladding in mixed α+β phase (c) does not account for azimuthal temperature variations. In order to overcome all these drawbacks of burst criterion, it is reasoned that artificial intelligence technique may be a better option to predict the burst parameters. Methods: Artificial neural network models based on feedforward backpropagation algorithm with logsig transfer function are developed. Results: Neural network architecture of 2-4-4-3, that is model with two hidden layers having four nodes in each layer is found to be the most suitable. The mean, maximum, and minimum prediction errors for this optimised model are 0.82%, 19.62%, and 0.004%, respectively. Conclusion: The burst stress, burst temperature, and burst strain obtained from burst criterion have average deviation of 19%, 12%, and 53% respectively whereas the developed neural network model predicted these parameters with average deviation of 6%, 2%, and 8%, respectively.

• 문화기술의 융합
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• 제10권3호
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• pp.299-304
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• 2024

미국과 소련이 맞서는 제1차 핵시대를 넘어 지금은 크고 작은 국가들로 핵 사용 결정권자가 다양화된 제2차 핵시대라고 할 수 있다. 이에 해당하는 국가인 인도와 파키스탄은 서로 적대국으로 맞서며 핵무기를 보유하고 있지만 핵태세, 핵전략은 상반돼 있다. 두 국가의 사례는 우리나라가 마주한 북한이 앞으로 어떤 핵태세를 취할 것인지에 대해 실마리를 줄 수 있다. 특히 파키스탄이 선택한 선행적 확전 태세는 상대 위협에 대해 핵무기를 선제적으로 쓸 수있다고 위협해서 적의 침략을 억제시키는 매우 공세적인 핵태세이다. 이는 선제공격할 수 있는 소규모 핵무기로도 할수 있는 옵션이다. 따라서 핵능력이 열세한 파키스탄이 인도의 위협에 대응하여 선택할 수 있는 최적의 태세로 보인다. 미국과 한국에 비해 열세인 북한은 앞으로도 파키스탄처럼 핵무기를 선제적으로 사용할 수 있다고 위협할 것으로 보인다. 반면 정권 유지를 위해 실제 사용하기까지는 인도와 같이 수세적이고 상당히 보수적일 것으로 전망된다.

• 한국태양에너지학회:학술대회논문집
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• 한국태양에너지학회 2011년도 춘계학술발표대회 논문집
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• pp.204-209
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• 2011

According to the environment report of UN, korea was classified as potable water shortage countries. Approximately 71% of the Earth's surface is covered by ocean. However, it is difficult to use for industry of residential purpose without a certain processing. The development of solar and waste-heat used absorption desalination technology have been examined as a viable option for supplying clean energy. In this study, the modelling of the main devices for solar and waste-heat used and adsorption desalination system was introduced. The design is divided into three parts. First, the evaporator for the vaporization of the top water is designed, and then the reactor for the adsorption and release of the steam is designed, followed by the condenser for the condensation of the fresh water is designed. In addition, new features based on the energy balance are also included to design absorption desalination system. In this basicresearch, One-bed(reactor) adsorption desalination plant that employ a low-temperature solar and waste energy was proposed and experimentally studied. The specific water yield is measured experimentally with respect to the time controlling parameters such as heat source temperatures, coolant temperatures, system switching and half-cycle operational times.

• 한국유체기계학회 논문집
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• 제16권4호
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• pp.35-43
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• 2013

Most of the nuclear power plants(NPPs) adopts pressure difference type flow meters such as venturi and orifice meters for the measurement of feedwater flow rates to calculate reactor thermal power. However, corrosion products in the feedwater deposits on the flow meter by fouling as operating time goes. These effects lead to severe errors in the flow indication and then determination of reactor thermal power. The averaging BDFT, which has developed by Yun et al., has a potentiality to minimize this problem thanks to its inherent measurement principle. Therefore, it is expected that the averaging BDFT can replace the venturi meter for the feedwater pipe of steam generator of NPPs. The present work compares the amplification factor, K, based on CFD calculation against the K obtained from experiments in order to confirm whether a commercial CFD code can be applicable to the evaluation of characteristic for the averaging BDFT. In addition to this, the simulations to take into account of fouling effect are also carried out by rough wall option. The results show that the averaging BDFT is a promising flow meter for the accurate measurement of flow rates in the fouling condition of the NPPs.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.192-205
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• 2020

Seismic design practices and seismic response analyses of civil structures and nuclear power plants (NPPs) have conventionally used the peak ground acceleration (PGA) or spectral acceleration (S_a) as an intensity measure (IM) of an earthquake. However, there are many other earthquake IMs that were proposed by various researchers. The aim of this study is to investigate the correlation between seismic responses of NPP components and 23 earthquake IMs and identify the best IMs for correlating with damage of NPP structures. Particularly, low- and high-frequency ground motion records are separately accounted in correlation analyses. An advanced power reactor NPP in Korea, APR1400, is selected for numerical analyses where containment and auxiliary buildings are modeled using SAP2000. Floor displacements and accelerations are monitored for the non- and base-isolated NPP structures while shear deformations of the base isolator are additionally monitored for the base-isolated NPP. A series of Pearson's correlation coefficients are calculated to recognize the correlation between each of the 23 earthquake IMs and responses of NPP structures. The numerical results demonstrate that there is a significant difference in the correlation between earthquake IMs and seismic responses of non-isolated NPP structures considering low- and high-frequency ground motion groups. Meanwhile, a trivial discrepancy of the correlation is observed in the case of the base-isolated NPP subjected to the two groups of ground motions. Moreover, a selection of PGA or S_a for seismic response analyses of NPP structures in the high-frequency seismic regions may not be the best option. Additionally, a set of fragility curves are thereafter developed for the base-isolated NPP based on the shear deformation of lead rubber bearing (LRB) with respect to the strongly correlated IMs. The results reveal that the probability of damage to the structure is higher for low-frequency earthquakes compared with that of high-frequency ground motions.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.427-438
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• 2013

In this paper we discuss spent fuel management options in the Republic of Korea (ROK) from two interrelated perspectives: Centralized dry cask storage and spent fuel pyroprocessing and burning in sodium fast reactors (SFRs). We argue that the ROK will run out of space for at-reactors spent fuel storage by about the year 2030 and will thus need to transition centralized dry cask storage. Pyroprocessing plant capacity, even if approved and successfully licensed and constructed by that time, will not suffice to handle all the spent fuel discharged annually. Hence centralized dry cask storage will be required even if the pyroprocessing option is successfully developed by 2030. Pyroprocessing is but an enabling technology on the path leading to fissile material recycling and burning in future SFRs. In this regard we discuss two SFR options under development in the U.S.: the Super Prism and the Travelling Wave Reactor (TWR). We note that the U.S. is further along in reactor development than the ROK. The ROK though has acquired more experience, recently in investigating fuel recycling options for SFRs. We thus call for two complementary joint R&D project to be conducted by U.S. and ROK scientists. One leading to the development of a demonstration centralized away-fromreactors spent fuel storage facility. The other involve further R&D on a combined SFR-fuel cycle complex based on the reactor and fuel cycle options discussed in the paper.