• Nuclear Engineering and Technology
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• 제39권3호
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• pp.171-186
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• 2007

Critical issues in the development targets for the future fast reactor(FR) cycle system, including sodium-cooled FR were to ensure safety assurance, efficient utilization of resources, reduction of environmental burden, assurance of nuclear non-proliferation, and economic competitiveness. A promising design concept of sodium-cooled fast reactor JSFR is proposed aiming at fully satisfaction of the development targets for the next generation nuclear energy system. A roadmap toward JSFR commercialization is described, to be followed up in a new framework of the Fast reactor Cycle Technology development(FaCT) Project launched in 2006.

• Nuclear Engineering and Technology
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• 제56권9호
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• pp.3950-3960
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• 2024

As a solution to the problem of spent nuclear fuels (SNFs), the disposal of SNF has gained attention from nations using nuclear energy because of hazards posed to the ecosystem. Among many proposed solutions, the most promising method is to dispose of SNF in a deep geological repository (DGR) which utilizes the multi-barrier concept developed by Finland and Sweden. Here, a new fully-coupled Thermal-Hydraulic (TH) code HADES (High-level rAdionuclide Disposal Evaluation Simulator) is developed using the MOOSE framework. This new code suggests basic numerical tools, such as a non-linear solver and finite element discretization, to assess the safety performance of disposal systems. The new TH code considered various TH behavior using Richards' flow approach, assuming gas pressure is constant. The HADES showed promising results when it was compared to various TH codes validated from DECOVAELX-THMC projects. When the single-canister model was utilized to estimate the TH behavior of the Korean Reference disposal System, although it showed significant saturation reduction due to the evaporation of water, the temperature was maintained under the thermal criteria limit, which is 100 ℃. In addition, the new code estimated temperature and degree of saturation of the multi-canisters model, considering two or three canisters, it showed a slightly lower temperature, 5 ℃, than the single-canister model. From these results, the following are concluded: (1) the new TH code contribute to an additional integrity by estimating TH behavior of KRS; (2) however, due to limitations in single-canister simulation, it is recommended to use multi-canisters simulation to estimate TH behavior accurately. Therefore, this model is anticipated not only to help licensing applications and estimation of various multi-physics phenomena and multi-canister at the disposal site.

• 에너지공학
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• 제11권4호
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• pp.359-369
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• 2002

원자력발전소는 여러 공학 분야의 기술이 결집된 거대 시스템이므로 연구개발의 우선 순위를 설정하는 기준을 정립하기가 간단하지 않고 설정 기준의 상대적 중요도에 대한 이해당사자간의 공감대가 부족하다. 본 연구는 이러한 배경에서 원전의 운영 관련 기술 체계를 정립하고 각 기술의 중요도 평가 기준을 설정하고 이 평가 기준들 간의 상대적 중요도를 파악하여 최종적으로는 각 기술의 상대적 중요도를 평가하였다. 연구 과정에서 기존의 연구기획 관련 문헌을 모두 조사하여 기술체계를 마련하고 계층화 분석법을 적용하여 기술간의 상대적 중요도를 결정하였으며 일관성지수와 현장인터뷰 결과를 활용하여 연구 결과의 검증을 시도하였다.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2952-2965
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• 2023

Advanced reactors, such as Small Modular Reactors or existing Nuclear Power Plants, often use Field Programmable Gate Array (FPGA) based controllers in new Instrumentation and Control (I&C) system architectures or as an alternative to existing analog-based I&C systems. Compared to CPU-based Programmable Logic Controllers (PLCs), FPGAs offer better overall performance. However, programming functions on FPGAs can be challenging due to the requirement for a hardware description language that does not explicitly support the operation of real numbers. This study aims to implement the Reactor Trip (RT) functions of the existing analog-based Reactor Protection System (RPS) using FPGAs. The RT equations for Overtemperature delta Temperature and Overpower delta Temperature involve dynamic compensators expressed with the Laplace transform variable, 's', which is not directly supported by FPGAs. To address this issue, the trip equations with the Laplace variable in the continuous-time domain are transformed to the discrete-time domain using the Z-transform. Additionally, a new operation based on a relative value for the equation range is introduced for the handling of real numbers in the RT functions. The proposed approach can be utilized for upgrading the existing analog-based RPS as well as digitalizing control systems in advanced reactor systems.

• 한국항행학회논문지
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• 제23권6호
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• pp.554-560
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• 2019

원전에서 각종 전자카드는 시간에 경과함에 따라 열화가 되므로 이에 대한 대책이 필요하다. 이 열화 카드들 중에서 노외중성자감시시스템의 카드들은 방사선원의 레벨에서 발생되는 중성자속을 총 원자로출력의 200%까지 연속적으로 감시하게 되는데, 원자로출력이 낮을 때의 경우와 높을 때의 경우의 출력감시신호처리 방법이 달라야 한다. 원자로 출력이 낮을 때는 대수적으로 발생되는 펄스신호를 선형적으로 계수하여 신호처처리를 해야 되지만, 원자로 출력이 커지게 되면 통계이론에 의한 방법으로 처리해야 정확한 값을 얻을 수 있기 때문이다. 이때 전자카드가 열화되는 것이 문제가 된다. 따라서, 본 연구에서는 저출력일 때와 고출력일 때의 신호처리 방법을 달리하여 일정한 기준에 의한 원자로의 출력레벨에서 이를 저출력에서 고출력으로 전환하기 위한 열화 입력선택회로를 개발하였다. 개발된 선택회로의 신뢰성을 확인하기 위하여 원전에서 사용되는 실제의 데이터값을 적용하여 테스트하였으며, 그 결과를 분석하여 선택회로의 정당성을 입증하였다.

• 한국항행학회논문지
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• 제24권6호
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• pp.557-565
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• 2020

원자로의 출력신호를 감시하는 노외중성자속감시계통의 열화상태를 점검하기 위해서는 원자로에서 방출되는 중성자 펄스를 감지하여 처리하는 전자카드에서 주파수형태로 감지하여 전압으로 변환한 후 대수 형태의 직류전압 값을 얻는 방법을 이용한다. 실제로 원전에서 적용하는 방법으로서는 주파수 카운터와 flip-flop 조합으로 이 과정을 수행하거나, 또는 다이오드펌프와 캐패시터의 조합을 이용하는 방법을 쓰며, 아직도 이 방법이 일반적으로 쓰이고 있다. 이 방법들은 높은 주파수에서는 신뢰성이 높으나 낮은 주파수에는 오차가 크고 측정시간도 오래 걸린다는 문제점이 있다. 따라서 본 연구에서는 고출력대의 고주파수 범위뿐만 아니라 중위출력 범위 주파수대, 그리고 극히 저출력 범위에 속해 있는 취약주파수대인 0.21 Hz~2 kHz 범위의 낮은 주파수대에 이르는 광범위한 주파수를 대수직류전압으로 신뢰성 높게 변환시킬 수 있는 장치를 개발하였다. 개발된 선택회로의 신뢰성을 확인하기 위하여 원전에서 사용되는 실제의 데이터값을 적용하여 테스트하였으며, 그 결과를 분석하여 선택회로의 정당성을 입증하였다.

• Bulletin of the Korean Chemical Society
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• 제7권2호
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• pp.89-96
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• 1986

The NMR chemical shift arising from 4d electron angular momentum and 4d electron spin dipolar-nuclear spin angular momentum interaction for a $4d^1$ system in a strong crystal field of octahedral symmetry, when the threefold axis is chosen as the quantization axis, has been investigated. A general expression using a nonmultipole expansion method is derived for the NMR chemical shift. From this expression all the multipolar terms are determined. We find that the nonmultipolar results for the NMR chemical shift ${\Delta}B$, is exactly in agreement with the multipolar results when $R {\ge} 0.20$ nm. It is also found that the 1/$R^7$ term contributes to the NMR chemical shift almost the same as the 1/$R^5$ in magnitude. The temperature dependence analysis of ${\Delta}B$/B(ppm) at various values of R shows that the 1/$T^2$ term has the dominant contribution to the NMR chemical shift but the contributions of other two terms are certainly significant for a $4d^1$ system in a strong crystal field of octahedral symmetry when the threefold axis is chosen to be the axis of quantization.

• 한국시스템다이내믹스연구
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• 제7권2호
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• pp.57-80
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• 2006

Nuclear technology made a great contribution to the national economy and society by localization of nuclear power plant design, and by stabilization of electricity price, etc. It is very important to conduct the retrospective analysis for the nuclear technology contribution to the national economy and society, but it is more important to conduct prospective analysis for the nuclear technology contribution. The term "technology value" is often used in the prospective analysis to value the result of technology development. There are various definitions of technology value, but generally it means the increment of future revenue or the reduction of future cost by technology development. These technology valuation methods are widely used in various fields (information technology or energy technology, etc). The main objective of this research is to develop valuation methodology that represents unique characteristics of nuclear power technology. The valuation methodology that incorporates market share changes of generation technologies was developed. The technology valuation model which consists of five modules (electricity demand forecast module, technology development module, market share module, electricity generation module, total cost module) to incorporate market share changes of generation technologies was developed. The nuclear power technology value assessed by this technology valuation model was 3 times more than the value assessed by the conventional method. So it was confirmed that it is very important to incorporates market share changes of generation technologies. The valuation results of nuclear power technology in this study can be used as policy data for ensuring the benefits of nuclear power R&D (Research and Development) investment.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.834-841
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• 2022

System thermal-hydraulic (STH) code is adopted for nuclear safety analysis. The critical flow model (CFM) is significant for the accuracy of STH simulation. To overcome the defects of current CFMs (low precision or long calculation time), a CFM based on a genetic neural network (GNN) has been developed in this work. To build a powerful model, besides the critical mass flux, the critical pressure and critical quality were also considered in this model, which was seldom considered before. Comparing with the traditional homogeneous equilibrium model (HEM) and the Moody model, the GNN model can predict the critical mass flux with a higher accuracy (approximately 80% of results are within the ±20% error limit); comparing with the Leung model and the Shannak model for critical pressure prediction, the GNN model achieved the best results (more than 80% prediction results within the ±20% error limit). For the critical quality, similar precision is achieved. The GNN-based CFM in this work is meaningful for the STH code CFM development.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.406-414
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• 2002

In TSPA of long-term post closure radiological safety on permanent disposal of HLW in Korea, appropriate management of input and output data through QA is necessary. The robust QA system is developed using the T2R3 principles applicable for five major steps in R&D＇s. The proposed system is implemented in the web-based system so that all participants in TSPA are able to access the system. In addition, the internet based input database for TSPA is developed. Currently data from literature surveys, domestic laboratory and field experiments as well as expert elicitation are applied for TSPA.