• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.425-426
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• 2005

1. The oxide thickness on the fuelled test rods was within the following range from 7 ${\mu}m$ to 17 ${\mu}m$. In general, the HANA claddings showed better corrosion behavior than the two reference alloys (A-Cladding and Zr-4). 2. The weight gains of corrosion coupons were ranged from 21 to 56 mg/$dm^2$.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4295-4306
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• 2023

The vibration of fuel rods in axial flow is a universally recognized issue within both engineering and academic communities due to its significant importance in ensuring structural safety. This paper aims to thoroughly investigate the stability and nonlinear vibration of a fuel rod subjected to axial flow in a newly designed high temperature gas cooled reactor. Considering the possible presence of thermal expansion and large deformation in practical scenarios, the thermal effect and geometric nonlinearity are modeled using the von Karman equation. By applying Hamilton's principle, we derive the comprehensive governing equation for this fluid-structure interaction system, which incorporates the quadratic nonlinear stiffness. To establish a connection between the fluid and structure aspects, we utilize the Galerkin method to solve the perturbation potential function, while employing mode expansion techniques associated with the structural analysis. Following convergence and validation analyses, we examine the stability of the structure under various conditions in detail, and also investigate the bifurcation behavior concerning the buckling amplitude and flow velocity. The findings from this research enhance the understanding of the underlying physics governing fuel rod behavior in axial flow under severe yet practical conditions, while providing valuable guidance for reactor design.

• Nuclear Engineering and Technology
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• 제24권3호
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• pp.228-235
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• 1992

처분된 폐기물에서 용해도가 큰 핵종이 침출될 때, 그 핵종의 용해도에 의해 조절되거나 조화 용해하지 않는 경우가 있다. 예를 들면 원자로 운영시 핵분열 생성물의 일부는 그레인 경계나 핵연료와 피복재 사이의 틈새에 축적될 수가 있다. 사용후 핵연료 처분장에서 이와 같이 축적된 핵분열 생성물중 세슘이나 요오드와 같이 용해도가 큰 핵종은 용기가 부식되면 지하수내에 급격하게 녹게된다. 이와 같이 틈새에 녹아있는 용해도가 큰 핵종의 이동현상을 시간 및 공간의 함수로 모사하고 그 수치 결과를 제시하였다. 전구간에서 유효한 근사해를 제시하고 이를 초기 및 후기 접근해와 Laplace 변환을 수치 재변환으로 얻은 해들과 비교함으로 검증하였다.

• Nuclear Engineering and Technology
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• 제30권4호
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• pp.328-341
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• 1998

This paper presents in detail the hydraulic characteristic measurements using LDV(Laser Doppler Velocimetry) in subchannels of a HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids and has a cylindrical configuration. Axial velocity and turbulent intensity were measured. The effects of the spacer grids on the turbulent flow were investigated using the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of the fuel bundle and the loss coefficients for the spacer grids were estimated from the measured pressure drops. The turbulent thermal mixing phenomena were discussed.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.899-906
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• 2018

The irradiation performance of $U_3Si$ dispersion fuel in an Al matrix, $U_3Si-Al$, under the Hi-Flux Advanced Neutron Application Reactor (HANARO) design full-power condition of 30 MW was tested for full-power qualification of the fuel. A test assembly was fabricated containing 18 fuel rods made with atomized $U_3Si$ powder manufactured at the Korea Atomic Energy Research Institute. The test assembly was irradiated for 188 full-power operation days in the HANARO subject to the normal fuel-loading scheme and achieved about 60 at% U-235 average burnup and 75 at% U-235 peak burnup. The maximum linear power of the test assembly was 98 kW/m. Nondestructive and destructive postirradiation examinations were conducted. The measured postirradiation examination data were compared with data from previous irradiations and the design criteria required for HANARO fuel. Consequently, it was concluded that in-pile performance was acceptable and fuel integrity was maintained, and the behavior satisfied the fuel design requirements.

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.7-13
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• 2012

A spacer grid which is one of the most important structural components in a pressurized water reactor fuel is an interconnected array of slotted grid straps, welded at the intersections to form an egg-crate structure. The spacer grid is required to not only protect fuel rods stably but also have sufficient lateral crush strength for the sake of enabling shut-down of the nuclear reactor during abnormal operating environments. Then, the lateral crush strength of the spacer grid is closely related with welding quality of the spacer grid. Previous research on the crush strength analysis of the spacer grid had been performed using only parent material properties. In this study, to investigate the effect on the crush strength of the spacer grid when used mechanical properties in weld zone instead of parent material properties, crush strength analysis considering mechanical properties in weld zone obtained from the instrumented indentation technique was performed and compared the results with the previous research.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.513-520
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• 2014

In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, $B_4C$, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations. The well-known WIMS-D4 and CITATION reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the $B_4C$; also the lowest PPF is gained using the Ag-In-Cd alloy. The maximum point power densities belong to the inside fuel regions surrounding the central flux trap (irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the graphite reflectors. The greatest and least fluctuation of the point power densities are gained by using $B_4C$ and Ag-In-Cd alloy, respectively.

• 대한기계학회논문집B
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• 제40권12호
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• pp.815-824
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• 2016

가압경수로에 장전되는 핵연료집합체는 연료 봉 다발과 지지격자 및 상하단 고정체로 구성되어 있다. 고온 고압의 냉각수는 원자로 하부로 유입되어 연료 봉 사이로 형성된 부수로를 따라 노심 상부로 흐른다. 경수로핵연료의 주요 열수력 성능인자는 정상운전시 압력강하 및 임계열속이며 사고시에는 급랭 시간이다. 한국원자력연구원에서는 경수로핵연료의 성능을 향상시키고 국산화를 위해 고성능 경수로핵연료, 이중냉각 핵연료 및 사고저항성 핵연료를 개발하였다. 경수로핵연료의 열수력 핵심기술을 개발하기 위해 압력강하 실험, 난류 유동혼합/열전달 실험, 임계열속 및 급랭 시험을 수행하였으며 전산유체역학 방법도 활용하였다. 더불어 사용후핵연료의 임시저장을 위한 건식저장 용기의 열유동에 대한 전산유체해석을 수행하였다. 한편, 경수로핵연료의 열수력 기반기술을 개발하고 실용화를 위해 대학 및 산업체와 협력연구도 진행하였다.

• 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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• 제10권1호
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• pp.26-36
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• 1978

Hot cell의 설계를 위하여 기사용 연료에서의 방사능과 붕괴 에너지의 수치적 계산을 하였다. 고리 1호기와 같은 경수로에서 거의 최대 연소율인 33,000MWD/T(U)으로 태워진 연료봉 시험을 위하여 보관할 수 있는 최적의 벽과 창 두께가 추정되었다. 기사용 연료를 hot cell에 넣기 전에 차폐물질의 두께 추정을 위해 그 연료를 여러 시간 간격동안 저장용기 속에서 냉각시켰다는 가정을 했다. 여러 종류의 차폐물질이 고려되었으며 방사선원과 관측점과의 거리도 변화시켜 보았다.