• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.590-592
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• 2008

A spacer grid is one of the most important structural components in a LWR nuclear fuel assembly. The primary considerations are to provide a Zircaloy spacer grid with crush strength sufficient to resist design basis loads, without significantly increasing pressure drop across the reactor core. In this study, the dynamic crush strength analysis and test are carried out for the specimens of a spacer grid assembly.

• 대한기계학회논문집A
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• 제35권3호
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• pp.281-287
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• 2011

얇은 관 탄성좌굴 공식의 불확실성을 고려하기 위해, 공식을 구성하는 파라미터인 튜브재료의 탄성계수, 푸아송 비, 튜브 두께 및 지름의 불확실성을 분석하였다. 본 연구는 원자로에서 연소되는 핵연료봉과 같이 사용 중 함몰을 엄격히 방지하고 있는 얇은 관의 설계신뢰도를 향상시키는 데에 중요하다. 분석 방법은 각각의 파라미터가 변화할 수 있는 범위를 충분히 포함할 수 있는 최소의 탄성좌굴 안전율을 구하고 이를 선형적으로 합하여 최종의 최소안전율을 구하였다. 최소 안전율에 가장 큰 영향을 미치는 파라미터는 관의 두께로 나타났다. 두께가 얇을수록 더 큰 최소안전율이 필요하며 예로 적용한 지르코늄 합금관의 경우, 두께가 0.254 와 0.87 mm 일 때 최소안전율은 각각 1.547 과 3.487 로 나타났다.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.327-362
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• 2005

The design, construction, and operation of a light water reactor (LWR) are subject to compliance with safety criteria specified for accident situations, such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA). Because reactor fuel is the primary source of radioactivity and heat generation, such a criterion is established on the basis of the characteristics and performance of fuel under the specific accident condition. As such, fuel behavior under accident situations impact many aspects of fuel design and power generation, and in an indirect manner, even spent fuel storage and management. This paper provides a comprehensive review of: the history of the current LOCA criteria, results of LOCA-related investigations on conventional and new classes of fuel, and status of on-going studies on high-burnup fuel under LOCA situations. The objective of the paper is to provide a better understanding of important issues and an insight helpful to establish new LOCA criteria for modem LWR fuels.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.489-498
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• 2011

UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2002년도 가을 학술발표회 논문집
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• pp.139-145
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• 2002

In this study, the geometric modeling has been conducted for 2 models of lower end fitting of advanced LWR fuel using three-dimensional solid modeler, Solidworks. Then, the optimization and the three-dimensional stress analysis using the finite element method has been peformed. The evaluation for the mechanical integrity of 2 models has been peformed based on the stress distribution obtained from the finite element analysis.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.811-820
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• 2013

SiC has been under investigation as a potential cladding for LWR fuel, due to its high melting point and drastically reduced chemical reactivity with liquid water, and steam at high temperatures. As SiC is a brittle material its behavior during the reflood phase of a Loss of Coolant Accident (LOCA) is another important aspect of SiC that must be examined as part of the feasibility assessment for its application to LWR fuel rods. In this study, an experimental assessment of thermal shock performance of a monolithic alpha phase SiC tube was conducted by quenching the material from high temperature (up to $1200^{\circ}C$) into room temperature water. Post-quenching assessment was carried out by a Scanning Electron Microscopy (SEM) image analysis to characterize fractures in the material. This paper assesses the effects of pre-existing pores on SiC cladding brittle fracture and crack development/propagation during the reflood phase. Proper extension of these guidelines to an SiC/SiC ceramic matrix composite (CMC) cladding design is discussed.

• KSTLE International Journal
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• 제10권1_2호
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• pp.27-32
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• 2009

In this study, fretting wear tests of nuclear fuel rods have been performed by using two kinds of spacer grid springs with a concave and a convex shape in room temperature dry and distilled water conditions. The objectives were to examine the variation of the wear mechanism with increasing fretting cycles and to evaluate the difference of the wear debris detachment behavior at each test environment. From the test results, the wear volume of each spring condition increased with increasing fretting cycles regardless of the test environments. However, the wear rate did not show a regular tendency and apparently changed with increasing fretting cycles. This is because the formation of the wear particle layer and/or the variation of the contact condition between the fuel rod and spring surfaces could affect a critical plastic deformation for detaching the wear debris. Based on the test results, the relationship between the wear behavior of each spring shape and test environment condition, and the variation of the surface characteristics are discussed in detail.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1156-1163
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• 2020

Fuel temperature coefficient (FTC) of PuO₂+ZrO₂ (ROX) fueled LWR cell is analyzed neutronically with reactor- and weapons-grade plutonium fuels in comparison with a U-free PuO₂+ThO₂ (TOX), and a conventional MOX fuel cells. The FTC value of a ROX fueled LWR is smaller compared to a TOX or a MOX fueled LWRs and becomes extremely positive especially, at EOL. This is because when fuel temperature is increased, thermal neutron spectrum is shifted to harder, which is extreme at EOL in ROX fuel than that in TOX and MOX fuels. Consequently at EOL, ²³⁹Pu and ²⁴¹Pu contributes to positive fuel temperature reactivity (FTR) in ROX fuel, while they have negative contribution in TOX and MOX fuels. The FTC problem of ROX fuel is mitigated by additive ThO₂, UO₂ or Er₂O₃. In ROX-additive fuel, the atomic density of fissile Pu becomes more than additive free ROX fuel especially at EOL, which is the main cause to improve the FTC problem. The density of fissile Pu is more effective to decrease the thermal spectrum shifts with increase the fuel temperature than additive ThO₂, UO₂ or Er₂O₃ in ROX fuel.

• 대한금속재료학회지
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• 제49권5호
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• pp.355-361
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• 2011

Creep behaviors of the Zr-1.Nb-0.5Cu (HANA-6) alloy strips with different orientations were investigated. Anisotropy was observed in the samples depending on their physical orientations due to the formation of texture in their microstructures. The creep strain rate was increased as the test stress and temperature increased. The rate was higher along the rolling-direction than in the transverse-direction irrespective of annealing conditions. However, the samples with $45^{\circ}$ direction showed different behaviors depending on the annealing temperature. When strips were finally annealed at $600^{\circ}C$ for 10 min, the primary creep rate of the $45^{\circ}$ strip was the highest among the various orientations although the saturated creep rate was the lowest. In the case of final annealing at $660^{\circ}C$ for 4 h, the highest creep rate occurred throughout the creep test in the $45^{\circ}$ strip. It is considered that the fraction of (100) planes along the direction of creep deformation affect the creep rates.

• Nuclear Engineering and Technology
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• 제15권1호
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• pp.57-69
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• 1983

경수형원자로 핵연료봉의 거동분석을 위한 전산코드인 FRAPCON-1 코드가 월성 1호기에 장전되는 CANDU형 핵연료봉의 거동분석을 위해 적절한지를 평가하였다. 연료내의 중성자속의 감소와 연료피복재간 열전달을 계산하는 FRAPCON-1 코드의 모형들을 수정하였으며 핵분열 기체방출모형의 CANDU 핵연료에 대한 타당성여부를 검토하였고 피복재와 냉각수간의 열전달 계수 계산을 위해 중수특성을 사용하였다. 수정된 코드 FRAPCON-1-CSK를 사용하여 월성 1호기 핵연료의 각 설계변수들에 대한 민감도 분석을 수행하였다. 아울러 월성 1호기 핵연료봉의 거동특성분석도 수행하였는데 계산된 결과들은 CANDU 핵연료봉에 대한 설계기준이 알러져 있지 않는 관계로 경수로 핵연료봉 설계기준의 입장에서 검토되었다.