• 대한기계학회논문집A
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• 제34권9호
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• pp.1175-1183
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• 2010

지지격자체는 경수로 핵연료집합체의 가장 중요한 핵심 구조부품 중에 하나이다. 질칼로이 지지격자체 설계시의 우선적으로 고려해야 할 사항은 지지격자체가 원자로심에서 냉각수의 심한 수두손실을 유발하지 않으면서 지진사고를 고려한 설계하중 하에서 충분한 횡방향 충격강도를 갖도록 하는 것이다. 본 연구에서는 시험과 유한요소해석을 통해 지지격자체의 횡방향 충격강도에 영향을 주는 인자들에 대한 분석을 수행하였고, 지지격자체 제조용 질칼로이 원자재 소요량을 획기적으로 줄이면서 지지격자체의 횡방향 충격강도를 개선할 수 있는 방안을 제시하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.93.1-93
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.113-113
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• 2004

• 방사성폐기물학회지
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• 제14권2호
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• pp.149-156
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• 2016

저장조에 위치한 사용후핵연료는 가혹한 원자로 조건에 의해 구조적 건전성이 와해되므로 외력에 취약하다. 따라서 운반 및 취급 중 사고 상황이 고려되어야 한다. 극단적인 경우, 핵연료 취급 중 사고로 인해 핵연료 저장조에서 핵연료집합체 낙하가 발생할 수 있다. 이러한 사고 상황 하에서 연료봉 파손 등을 평가하기 위해서 수조에 충돌할 때 발생하는 충돌력을 분석할 필요가 있다. 이는 핵연료가 수조 바닥에 충돌할 때의 속도를 입력으로 하여 평가될 수 있다. 연료봉이 핵연료 중량 및 부피의 대부분을 차지하고 있으므로, 연료봉 다발은 수중 항력을 예측하는데 중요한 역할을 한다고 볼 수 있다. 본 연구에서는 $3{\times}3$ 의 짧은 연료봉 다발을 모델로 사용하여 수중에서 낙하할 때 받는 수력을 계산하였고, 이를 전산모사와의 비교를 통하여 검증하였다. 본 방법론을 사용후핵연료 건전성 평가에 적용할 수 있을 것으로 기대된다.

• 설비공학논문집
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• 제29권10호
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• pp.538-550
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• 2017

In a PWR (Pressurized Water Reactor), the appropriate heat removal from the surface of fuel rod bundle is important for ensuring thermal margins and safety. Although many CFD (Computational Fluid Dynamics) software have been used to predict complex flows inside fuel assemblies with mixing vanes, there is no domestic regulatory guideline for the comprehensive evaluation of CFD software. Therefore, from the nuclear regulatory perspective, it is necessary to perform the systematic assessment and prepare the domestic regulatory guideline for checking whether valid CFD software is used for nuclear safety problems. In this study, to provide systematic evaluation and guidance on the applicability of CFD software to the domestic nuclear safety area, the results of the sensitivity analysis for the effect of the discretization scheme accuracy for the convection terms and turbulence models, which are main factors that contribute to the uncertainty in the calculation of the nuclear safety problems, on the prediction performance for the turbulent flow distribution inside the fuel assembly with split-type mixing vanes were explained.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3709-3715
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• 2023

It is an important work to measure the dimensions of underwater spent fuel assemblies in the nuclear power industry during the overhaul, to judging whether the spent fuel assemblies can continue to be used. In this paper, a three dimensional reconstruction method for underwater spent fuel assemblies of nuclear reactor based on linear structured light is proposed, and the topography and size measurement was carried out based on the reconstructed 3D model. Multiple linear structured light sensors are used to obtain contour size data, and the shape data of the whole spent fuel assembly can be collected by one-dimensional scanning motion. In this paper, we also presented a corrected model to correct the measurement error introduced by lead-glass and water is corrected. Then, we set up an underwater measurement system for spent fuel assembly based on this method. Finally, an underwater measurement experiment is carried out to verify the 3D reconstruction ability and measurement ability of the system, and the measurement error is less than ±0.05 mm.

• Tribology and Lubricants
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• 제30권6호
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• pp.343-349
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• 2014

Currently, in order to resist the scratching of the fuel rod surface while fabricating the fuel assembly of the light-water nuclear reactor, we use a solution of nitrocellulose, an explosive material, as a dry lubricant along with its solvent. However, the demand for developing safe and harmless aqueous alternative materials for environment-conservation and field-worker safety has increased. In this study, we demonstrate the preparation of a novel aqueous resin composite using a formulation of aqueous polymeric resin, alcoholic solvent, and water. Subsequently, we characterize this composite on the basis of hardness, adhesive property, and water solubility using plates similar to the fuel rod material. The insertion test of a fuel rod coated with the YS-3 composite shows load values of $18.8-20.5kg/cm^2$, which is comparable with $18.8-20.5kg/cm^2$ of the nitrocellulose coating agent. In addition, the depth and width of longitudinal scratches caused by the YS-3 composite test are 50% higher than those of the standard. We can develop a harmless and safe aqueous dry lubricant to replace the existing NC products through field testing of 264 pieces of fuel rods, after producing 350 kg of the YS-3 prototype. The scratch test for the rod surface showed that weight of chip of YS-3 prototype was smaller than that of NC before and after solvent treatment, indicating the properties of YS-3 prototype was comparable to the counterpart.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2003년도 추계학술발표대회 개요집
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• pp.55-57
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• 2003

The fuel assemblies as the nuclear fuel for the pressurized water reactor(PWR) are loaded in the reactor core throughout the residence time of three to five years. The fuel assembly is manufactured using special welding processes and under strict quality assurance and control systems. In this paper welding parts, welding methods, and welding tests for the integrity of the PWR fuel assemblies are introduced.

• 한국소음진동공학회논문집
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• 제16권2호
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• pp.163-168
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• 2006

The PWR Nuclear Fuel assembly consists of more than 250 fuel rods that are supported by leaf springs in the cells of more than 10 Spacer Grids (SG) along the rod length. Since it is not easy to conduct mechanical tests on a full-scale model basis, the small-scaled rod bundle $(5\times5)$ which is called partial fuel assembly is generally used for various performance tests during the development stage. As one of the small-scaled tests, a flow test should be carried out in order to verify the performance of the spacer grid to obtain the Flow-Induced Vibration (FIV) characteristics of the scaled fuel assembly over the specified flow range. A vibration test should be also performed to obtain the modal parameters of the assembly prior to the flow test. In this study, we want to develop the estimation procedure of the damping ratio for the scaled test assembly. For the damping factor of the partial fuel assembly and the grid cage at the first vibration mode, as one of the vibration tests, a so-called pluck testing has been performed in air as a preliminary test prior to in-flow damping measurement test. Logarithmic decrement method is used for calculation of the damping ratio. Estimated damping ratio of the partial fuel assembly is about $0.7\%$ with reasonable error of $2\%$ for the previous results. Nonlinear behavior of the partial fuel assembly might be stem mainly from the rod-grid support configuration.

• Advances in Energy Research
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• 제5권2호
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.