• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1629-1635
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• 2017

Reliable prediction of nuclear fuel rod behavior of nuclear power reactors constitutes a basic demand for steady-state calculations, design purposes, and fuel performance assessment. Perfect design of fuel rods as the first barrier against fission product release is very important. Simulation of fuel rod performance with a code or software is one of the fuel rod design steps. In this study, a software program called MARCODE is developed in MATLAB environment that can analyze the temperature distribution, gap conductance value, and fuel and clad displacement in both solid and annular fuel rods. With a comparison of the maximum fuel temperature, fuel average temperature, fuel surface temperature, and gap conductance in solid and annular fuel, the effects of a central hole on the fuel temperature distribution are investigated.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2000년도 춘계학술대회논문집
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• pp.639-646
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• 2000

Estimation for the vibration behavior of a nuclear fuel rod with its supporting structure called spacer grid has been made by the both of experimental and analytical methods in order to compare the supporting performance of two kinds of the spacer grids which have been newly developed. For the analytical method the fuel rod was modeled as a beam continuously supported by the springs of the spacer grid, and ABAQUS computer code was utilized. After a modal testing was performed for the fuel rod supported by five spacer grids, two results has been compared to justify and compensate the both methods. It has been found that the spring design of the spacer grid could give significant effect to natural frequency and vibration amplitude of the fuel rod.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2007년도 춘계학술대회논문집
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• pp.197-201
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• 2007

Vibration characteristics of an annular cylindrical fuel rod, which was proposed as a candidate design of fuel's cross section for the ultra-high burnup nuclear fuel, according to the cross-sectional dimensions and the number of supports or the span length were analytically studied. Finite element(FE) modeling for the annular cross sectional fuel was based on the methodology, that have been proven by the test verification, for the conventional PWR nuclear fuel rod. A commercial FEA code, ABAQUS, was used for the FE modeling and analysis. A planar beam element (B21) that uses a linear interpolation was used for the fuel rod and a linear spring element for the spring and dimple of the SG. Natural frequencies and mode shape were calculated according to the preliminary design candidates for the fuel's cross sectional dimension and the number of span. From the analysis results, the design scheme of the annular fuel compatible to the present PWR nuclear reactor core was discussed in terms of the number of supports and fuel's cross section.

• 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.

• 대한전자공학회논문지SP
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• 제47권1호
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• pp.91-96
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• 2010

본 논문에서는 핵연료봉의 변형에 대한 고정도 검사방법을 제안한다. 핵 연료봉과 이를 관측하는 영상 센서의 광축을 수직으로 구성한다. 영상 센서의 광축을 기준으로 45도 또는 그보다 높은 각도로 레이저 라인빔을 연료봉 표면에 조사하면 연료봉의 수평 방향 변위가 영상 센서에서는 수직 방향 변위로 관측된다. 핵 연료봉 표면에 일정 각도로 입사된 레이저 라인빔이 영상 센서면에서는 일정 두께를 갖는 포물선 형태로 관측되게 된다. 센서 화면에 나타나는 일정 두께의 포물선을 영상처리하여 타원으로 모델링하고 타원의 장축과 단축의 기울기를 구한다. 포물선의 변곡점과 모델링한 타원의 장축과 단축이 교차하는 지점을 특징점으로 추출한다. 이와 같은 영상처리 알고리즘을 이용하여 핵 연료봉의 수평방향 변위에 따른 특징점 좌표의 수직방향 편차를 계산한다. 크러드가 형성된 핵연료봉 시편에 대해 고해상도 영상센서를 사용하여 실험한 결과 중성자 조사후 핵연료봉의 변형 검사기준인 $150{\mu}m$ 보다 3배 이상 개선된 $50{\mu}m$ 이하의 검사 정밀도를 달성하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.680-686
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• 1995

It has been known that water-side corrosion of fuel rods in nuclear reactor is accompanied with the loss of metallic wall thickness and pickup of hydrogen. This corrosion is one of the important limiting factors ill the operating life of fuel rods. In connection with the fuel cladding corrosion, a device to measure the water-side oxide layer thickness by means of the eddy-current method without destructing the fuel rod was developed by KAERI. The device was installed on the multi-function testing bench in the nondestructive test hot-cell and its calibration was carried out successfully for the standard rod attached with plastic thin films whose thicknesses are predetermined. It shows good precision within about 10% error. And a PWR fuel rod, one of the J-44 assembly discharged from Kori nuclear power plant Unit-2, has been selected for oxide layer thickness measurements. With the result of data analysis, it appeared that the oxide layer thicknesses of Zircaloy cladding vary with the length of the fuel rod, and their thicknesses were compared with those of the destructive test results to confirm the real thicknesses.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1333-1344
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• 2019

An integrated approach of model simplification for high burnup spent nuclear fuel is proposed based on material calibration using optimization. The spent fuel rods are simplified into a beam with a homogenous isotropic material. The proposed approach of model simplification is applied to fuel rods with two kinds of interfacial configurations between the fuel pellets and cladding. The differences among the generated models and the effects of interfacial bonding efficiency are discussed. The strategy of model simplification adopted in this work is to force the simplified beam model of spent fuel rods to possess the same compliance and failure characteristics under critical loads as those that result in the failure of detailed fuel rod models. It is envisioned that the simplified model would enable the assessment of fuel rod failure through an assembly-level analysis, without resorting to a refined model for an individual fuel rod. The effective material properties of the simplified beam model were successfully identified using the integrated optimization process. The feasibility of using the developed simplified beam models in dynamic impact simulations for a horizontal drop condition is examined, and discussions are provided.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1987년도 한국자동제어학술회의논문집; 한국과학기술대학, 충남; 16-17 Oct. 1987
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• pp.613-618
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• 1987

Rod consolidation is a method of increasing spent nuclear fuel storage capacity by disassembling fuel assemblies thus storing the fuel rods in a tighter array. It involves some basic operations which closely resemble to the material handling of a manufacturing process. But all the operations must be controlled remotely in shielded environment from outside due to the highly radioactive nature of the workpiece. In this study the status of the rod consolidation technology in other countries has been surveyed and a feasibility study for the conceptual design of this process have been made.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2003년도 춘계학술대회 논문집
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• pp.1154-1157
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• 2003

The spent fuel slitting device is an equipment developed for the separation of the pellet and hull from the cutting fuel rod with length of 250 mm, and in order to feed UO$_2$ pellet. We have analyzed on the existing technologies for designing and producing of the slitting device in the first year(2001), based on these results, designed and produced the rod slitting device. It has effectively separated the pellet from the hull, but demanded the supplement separation work because of the mixing with pellet and hull in the vessel, and required the condition for the reducing time of the process. In the second year(2002), we have reduced the work time, performed the test and capacity evaluation with the improving device, based these results, and ensured the data demanded for designing of the spent fuel rod slitting device. We have compared with the DUPIC(Direct use of spent PWR fuel in CAND reactors) process, and developed the device for the purpose of reducing over 40 % in comparition with the DUPIC operation time(5 minutes). Based on these results, it will is effectively applied to available data for designing and producing of the hot test facility.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3704-3710
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• 2021

The characteristics of the external radiation on the surface of the casks for spent nuclear fuel (SNF) storage by dry method are investigated for the case when the spatial distribution of SNF in the basket changes due to the destruction of the fuel rod claddings. The surface areas are determined, where the changes in fluxes of neutrons, produced by ²⁴⁴Cm actinide, and γ-quanta, produced by long-lived isotopes, are maximum in the result of the decrease in the height of the SNF area. Concrete (VSC-24) and metal (SC-21) casks are considered as examples. The procedure of periodic measurement of the dose rate of neutrons or γ-quanta at the specified points of the cask surface is proposed for identifying the fuel rod cladding destruction. Under normal operation, the decrease in the dose rate produced by neutrons as the function of SNF storage duration is determined by the half-life of ²⁴⁴Cm, and for γ-quanta - by the half-lives of long-lived SNF isotopes. Consequently, a stepwise change in the dose rate of neutrons or γ-quanta, detected by the measurements, as compared to the previous one, would indicate the destruction of the fuel rod claddings.