• Nuclear Engineering and Technology
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• 제44권2호
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• pp.177-198
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• 2012

This paper reviews the fast reactor physics and computational methods. The basic reactor physics specific to fast spectrum reactors are briefly reviewed, focused on fissile material breeding and actinide burning. Design implications and reactivity feedback characteristics are compared between breeder and burner reactors. Some discussions are given to the distinct nuclear characteristics of fast reactors that make the assumptions employed in traditional LWR analysis methods not applicable. Reactor physics analysis codes used for the modeling of fast reactor designs in the U.S. are reviewed. This review covers cross-section generation capabilities, whole-core deterministic (diffusion and transport) and Monte Carlo calculation tools, depletion and fuel cycle analysis codes, perturbation theory codes for reactivity coefficient calculation and cross section sensitivity analysis, and uncertainty analysis codes.

• Nuclear Engineering and Technology
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• 제26권1호
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• pp.119-125
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• 1994

핵연료봉의 과도상태 출력준위는 핵연료봉의 과도상태 거동에서 가장 중요한 변수중의 하나이다. 핵연료 성능 데이타베이스의 분석과 핵연료의 과도상태 거동에 영향을 줄 수 있는 핵연료봉 출력이력, 속중성자속, 농축도 및 주기길이 등의 인자들의 민감도 분석을 통해서, ANS 과도조건 I 및 II에서 핵연료봉의 기계적 건전성이 유지되는 허용가능 과도상태 출력을 구하기 위해 일반적으로 적용이 가능한 방법론이 유도되었으며, 이를 통해 17$\times$17 KOFA 핵연료봉의 허용가능 과도상태 출력이 연소도의 함수로써 결정되었다. 이 방법론을 도입함으로써, 현재와 같이 매 주기마다 핵연료봉 과도상태 설계분석을 수행할 필요가 없이 단지 해당주기에서의 과도상태 최대 출력준위 평가로써 핵연료봉의 과도상태 설계를 대체할 수 있으며, 17$\times$17 KOFA 핵연료에 대하 낮은 연소도영역에서 기존의 최대 허용 과도상태 출력 준위인 591 w/cm보다 큰 최대 689.5 w/cm까지 허용함으로써 원자로 운전에 유연성을 줄 수 있다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2009년도 학술논문요약집
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• pp.55-56
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• 2009

From the OT cycle analysis results, the nuclear power demand grows to ~70 GWe in 2150. The SF and TRU out-core inventories in 2150 will be 186500 t and 2100 t, respectively. The MOX fuel cycle gives 84% and 9% lower values for the SF and out-core TRU inventories, respectively.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.624-634
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• 2016

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

• 방사성폐기물학회지
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• 제6권2호
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• pp.147-154
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• 2008

국내 원자력발전소의 사용후핵연료 저장용량의 포화가 10년 이내에 예상되고 있으며, 소외중간저장방식을 관리 방안으로서 선정할 경우, 상당한 양의 사용후핵연료 운반이 해상 혹은 육로를 통해 매년 이루어져야만 한다. 본 논문에서는 4곳의 원자력발전소부지에 분산 저장되어 있는 사용후핵연료를 해안에 위치한 가상의 중간저장시설과 영구처분시설을 대상으로 사용후핵연료 운반물량을 효과적으로 분석할 수 있는 체계를 구축하였다. 각 발전소 부지, 중간저장시설, 영구처분시설의 저장고를 중심으로 사용후핵연료 물질 수지식을 세우고, 이에 대한 해를 VISUAL BASIC으로 구하여 운반 물량 분석이 용이하게 수행할 수 있는 컴퓨터 프로그램(CASK)을 개발하였다. 개발된 물량 분석 프로그램을 활용하여 4개 원자력 발전소 부지로부터 사용후핵연료 운반 물량을 분석하고, 운반물량 파라미터 분석을 통하여 본 프로그램의 활용도를 보였다. 개발된 사용후핵연료 운반 물량 분석 체계는 향후 운반비용 분석에도 유용하게 활용될 것이다.

• 방사성폐기물학회지
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• 제20권4호
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• pp.469-488
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• 2022

Technology for high-level-waste disposal employing a multibarrier concept using engineered and natural barrier in stable bedrock at 300-1,000 m depth is being commercialized as a safe, long-term isolation method for high-level waste, including spent nuclear fuel. Managing heat generated from waste is important for improving disposal efficiency; thus, research on efficient heat management is required. In this study, thermal management methods to maximize disposal efficiency in terms of the disposal area required were developed. They efficiently use the land in an environment, such as Korea, where the land area is small and the amount of waste is large. The thermal effects of engineered barriers and natural barriers in a high-level waste disposal repository were analyzed. The research status of thermal management for the main bedrocks of the repository, such as crystalline, clay, salt, and other rocks, were reviewed. Based on a characteristics analysis of various heat management approaches, the spent nuclear fuel cooling time, buffer bentonite thermal conductivity, and disposal container size were chosen as efficient heat management methods applicable in Korea. For each method, thermal analyses of the disposal repository were performed. Based on the results, the disposal efficiency was evaluated preliminarily. Necessary future research is suggested.

• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1181-1190
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• 2009

Knowing more about the Local Power Density (LPD) at the hottest part of a nuclear reactor core can provide more important information than knowledge of the LPD at any other position. The LPD at the hottest part needs to be estimated accurately in order to prevent the fuel rod from melting in a nuclear reactor. Support Vector Machines (SVMs) have successfully been applied in classification and regression problems. Therefore, in this paper, the power peaking factor, which is defined as the highest LPD to the average power density in a reactor core, was estimated by SVMs which use numerous measured signals of the reactor coolant system. The SVM models were developed by using a training data set and validated by an independent test data set. The SVM models' uncertainty was analyzed by using 100 sampled training data sets and verification data sets. The prediction intervals were very small, which means that the predicted values were very accurate. The predicted values were then applied to the first fuel cycle of the Yonggwang Nuclear Power Plant Unit 3. The root mean squared error was approximately 0.15%, which is accurate enough for use in LPD monitoring and for core protection that uses LPD estimation.

• Asian Journal of Atmospheric Environment
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• 제4권1호
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• pp.20-25
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• 2010

The gaseous $^{222}Rn$ concentration at the level of clouds was estimated by using the wet scavenging model of its decay products with the observed data of environmental radiation at the ground. And the origin of the $^{222}Rn$ was also discussed. The estimation was done for a precipitation event on Dec. 26-27, 2003, when a large increase of the radiation was observed in Tokai-mura in Ibaraki, Japan. From a backward trajectory analysis, the origin of $^{222}Rn$ atoms for that event was traced back to the northeastern part of China, and it was expected that the large amount of $^{222}Rn$ emitted in the northeastern part of China was transported to Tokai-mura by the Eurasian continental air mass.

• 방사성폐기물학회지
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• 제5권3호
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• pp.229-238
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• 2007

본 논문에서는 고준위폐기물 처분용기를 지하 심지층에 처분하기 위하여 요구되는 구조설계 요구조건과 구조안전성 평가 기준을 도출하였다. 고준위폐기물은 높은 열과 많은 방사능을 방출하기 때문에 고준위폐기물을 넣어 보관하는 처분용기는 그 취급에 많은 주의가 요구된다. 이를 위하여 고준위폐기물 처분용기는 장기간(보통 10,000년 동안) 안전한 장소에 보관되어야 한다. 보통 이 보관 장소는 지하 500m에 위치한다. 지하 깊은 화강암에 고준위폐기물을 보관하도록 설계되는 처분용기는 내부주철삽입물과 이를 감싸고 있는 부식에 강한 와곽쉘, 위 덮개와 아래 덮개로 구성되는 구조로 되어 있으며 지하수압과 벤토나이트 버퍼의 팽윤압을 받는다. 따라서 고준위폐기물 처분용기는 심지층에 보관 시 이들 외력들을 견디도록 설계되어야 한다. 만약에 발생 가능한 모든 하중조합을 고려한 처분용기 설계가 되지 않으면 심지층에 위험한 고준위폐기물 처분 시에 처분용기에 소성변형이나 크랙 또 좌굴같은 구조적 결함이 발생할 수 있다. 따라서 심지층에 처분용기를 처분 시에 처분용기에 발생하는 구조적 문제들이 발생하지 않게 하기 위하여 여러 가지 구조해석이 수행되어야 한다. 이러한 구조해석 수행에 앞서 처분용기 설계 타당성을 평가하기 위한 기준이 필요하다. 또한 평가기준에 영향을 미치는 설계요구조건(설계변수)이 명확히 검토되어야 한다. 따라서 본 논문에서는 처분용기의 구조설계 요구조건(설계변수)과 구조 안전성 평가기준을 도출하고자 한다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.21-26
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• 1997

Recently, thorium is receiving increasing attention as an important fertile material for the expanding nuclear power programs around the world. The superior nuclear and physical properties of thorium-based fuels could lead to very low fuel cycle cost and make thorium reactors economically attractive. In addition, the use of thorium in reactors would permit more efficient utilization of low cost uranium reserves and reduction nuclear wastes. In this work, the nuclear characteristics of a new type thorium fueled reactor (Radkowsky Thorium Reactor) consisting seed-blanket assemblies are addressed and compared with those typical assemblies of a PWR (CE type). Also, an assessment on several advantages of thorium fueled reactors is provided. All these results are based on the HELIOS code calculation.