• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.545-546
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• 2018

• 비파괴검사학회지
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• 제18권4호
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• pp.321-329
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• 1998

발전설비의 건전성은 원자력발전소의 안전운영을 위하여 필수적이며 이를 위하여 원자력발전소의 안전 규제법에서는 원전사업자 책임하에 신뢰성있는 가동중검사를 주기적으로 수행하도록 규정하고 있다. 따라서 한전은 국내 유일의 원전사업자로서 가동중검사의 정확도 및 신뢰성을 높이기 위한 기술개발에 최선을 다하고 있으며, 전력연구원은 한전의 기술개발 및 연구업무의 중추적 역할을 담당하는 부서로서 효율적인 원전 가동중검사 수행을 위하여 종합지원체계 (total support system : TSS)를 확립하고, 국내기술자립을 위한 요소기숱개밭은 물론 검사요원의 교육 및 자격인증체계 확립에 만전을 기하고 있다. 이 논문은 이와 같은 원자력발전소의 발전설비 건전성 진단을 위하여 현재 전력연구원에서 수행중인 가동중검사 기술개발업무 전반에 관해 상세하게 설명하기 위한 것이다.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.655-662
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• 2007

Most decision makers in the electricity industry plan their electric power expansion program by considering only a least cost operation, even when circumstances change with differing complexities. It is necessary, however, to analyze a long-term power expansion plan from various points of view, such as environmental friendliness, benefit of a carbon reduction, and system reliability, as well as least cost operation. The objective and approach of this study is to analyze the proper role of nuclear power in a long-term expansion plan by comparing different scenarios in terms of the system cost changes, $CO_2$ emission reduction, and system reliability in relation to the Business-As-Usual (BAU). The conclusion of this paper makes it clear that the Korean government cannot but expand the nationwide nuclear power program, because an increased energy demand is inevitable and other energy resources will not provide an adequate solution from an economic and sustainability point of view. The results of this analysis will help the Korean government in its long-term resource planning of what kinds of role each electric resource can play in terms of a triangular dilemma involving economics, environmental friendliness, and a stable supply of electricity.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.35-42
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• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.929-940
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• 2013

Environmental fatigue of the metallic components in light water reactors has been the subject of extensive research and regulatory interest in Korea and abroad. Especially, it was one of the key domestic issues for the license renewal of operating reactors and licensing of advanced reactors during the early 2000s. To deal with the environmental fatigue issue domestically, a systematic test program has been initiated and is still underway. The materials tested were SA508 Gr.1a low alloy steels, 316LN stainless steels, cast stainless steels, and an Alloy 690 and 52M weld. Through tests and subsequent analysis, the mechanisms of reduced low cycle fatigue life have been investigated for those alloys. In addition, the effects of temperature, dissolved oxygen level, and dissolved hydrogen level on low cycle fatigue behaviors have been investigated. In this paper, the test results and key analysis results are briefly summarized. Finally, an on-going test program for hot-bending of 347 stainless steel is introduced.

• 방사성폐기물학회지
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• 제19권2호
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• pp.243-253
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• 2021

In the dismantling process of a reactor coolant system (RCS) piping, a radiation protection plan should be established to minimize the radiation exposure doses of dismantling workers. Hence, it is necessary to estimate the individual effective dose in the RCS piping dismantling process when decommissioning a nuclear power plant. In this study, the radiation exposure doses of the dismantling workers at different positions was estimated using the MicroShield dose assessment program based on the NUREG/CR-1595 report. The individual effective dose, which is the sum of the effective dose to each tissue considering the working time, was used to estimate the radiation exposure dose. The estimations of the simulation results for all RCS piping dismantling tasks satisfied the dose limits prescribed by the ICRP-60 report. In dismantling the RCS piping of the Kori-1 or Wolsong-1 units in South Korea, the estimation and reduction method for the radiation exposure dose, and the simulated results of this study can be used to implement the radiation safety for optimal dismantling by providing information on the radiation exposure doses of the dismantling workers.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2835-2843
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• 2023

After an emergency shutdown of a lead-bismuth fast reactor, thermal stratification occurs in the upper Plenum, which negatively impacts the integrity of the reactor structure and the residual heat removal capacity of natural circulation flow. The research on thermal stratification of reactors has mainly been conducted using an experimental method, a system program, and computational fluid dynamics (CFD). However, the equipment required for the experimental method is expensive, accuracy of the system program is unpredictable, and resources and time required for the CFD approach are extensive. To overcome the defects of thermal stratification analysis, a high-precision full-order thermal stratification model based on CFD technology is prepared in this study. Furthermore, a reduced-order model has been developed by combining proper orthogonal decomposition (POD) with Galerkin projection. A comparative analysis of thermal stratification with the proposed full-order model reveals that the reduced-order thermal stratification model can well simulate the temperature distribution in the upper plenum and rapidly elucidate the thermal stratification interface characteristics during the lead-bismuth fast reactor accident. Overall, this study provides an analytical tool for determining the thermal stratification mechanism and reducing thermal stratification.

• 기술혁신학회지
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• 제12권1호
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• pp.70-87
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• 2009

최근 유가급등, 실물경제 위기 및 에너지 환경 위기 시대에 직면하여 공공부문에서의 효율성 증대에 대한 요구는 높아지고 있다. 그간 국가연구개발사업의 효율성은 개별 사업별 절대적 성과치를 근거로 재량적으로 판단되어 온 경우가 많은 점을 인식하고, 본 연구는 사업별 비교를 통해 원자력연구개발사업의 상대적 효율성을 도출함으로써 객관적 결과를 도출함을 목적으로 하였다. 본 연구를 수행하기 위해 교육과학기술부 주요 사업인 기초연구개발사업, 특정연구개발사업 및 원자력연구개발사업의 성과보고서에 제시된 자료를 활용하였고, 문헌검토와 관련이론을 통해 분석방법 및 모형을 설계하였다. 한편 상대적 효율성을 판단하기 위하여 자료포락분석법을 활용하였으며 이의 분석결과를 토대로 원자력연구개발사업의 효율성 증진방안을 세부사업별로 제시하였다. 연구개발사업의 효율성 측정을 위하여 자료포락분석법이 활용된 사례가 없는 만큼, 동 분석법의 한계를 극복할 수 있는 개선된 자료포락분석법이 적용된다면 연구개발사업의 효율적 관리에 기여할 것으로 기대한다.

• Journal of Radiation Protection and Research
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• 제27권2호
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• pp.111-120
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• 2002

IAEA가 제시한 중 저준위 방사성폐기물 천층처분시설 기준 안전성평가 사례에 대해 MASCOT 프로그램을 이용하여 안전성평가를 수행하였다. 이를 위해 기준시나리오에 대한 개념 모델을 개발하였다. 지질계와 생태계의 연결매체인 우물을 동한 지하수 이동경로에 대한 평가를 수행하였고 생태계 모델에서는 구획모델을 적용하여 인간활동을 통한 최종 방사선적 영향을 평가하였으며, 다른 평가 결과와의 비교를 통해 기준시나리오에 대한 개념모델의 적합성을 조사하였다. 본 연구 결과는 구획모델을 이용한 지하수 유동경로에 대한 대표적인 개념모델을 총체적인 처분시스템의 안전성평가에 만족스럽게 이용할 수 있다는 것을 보여주었다. 또한 MASCOT 프로그램을 이용하여 복잡하고 다양한 이동경로를 통한 천층처분시설의 방사선적 안전성평가가 가능함을 보였다.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1349-1359
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• 2016

Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX ($T_{co}$) and the hot outlet temperature of the intermediate heat exchanger ($T_{ho2}$) by manipulating the hot-side flow rates of the heat exchangers ($F_h/F_{h2}$) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX ($T_{co}$) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.