• 한국소음진동공학회논문집
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• 제17권9호
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• pp.819-825
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• 2007

Vibration characteristics of an duel-cooling 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.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3095-3105
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• 2022

Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7×7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9×9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.

• 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.

• Nuclear Engineering and Technology
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• 제25권3호
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• pp.424-429
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• 1993

핵연료 소결체가 장전되는 핵연료봉의 끝부분에는 각각의 핵연료봉을 구분해주는 고유의 문자가 인쇄되어 있다. 핵연료 집합체 제조 과정에서 각각의 핵연료봉은 고유 문자에 의해 구분되어 체계적으로 관리되고 있으며 아울러 핵연료 연소 이상상태 감시 및 사용후 핵연료 검사 분야에서 핵연료봉 제조과정 추적에 이용되고 있다. 핵연료봉 문자 자동인식은 핵연료 집합체 제조과정의 자동화를 위한 핵심 기술이다. 본 연구에서는 핵연료봉 문자인식 시스템을 개발하여, 핵 연료봉단에 기록된 각 문자로 부터 추출한 메쉬 특징값을 데이타베이스에 저장된 특정 문자의 특징값과 비교하여 자동으로 문자인식을 수행하도록 하였다. 실험 결과, 95.83 퍼센트의 양호한 인식률을 기록하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.1859-1863
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• 2007

A new type of a fretting wear tester has been designed and developed in order to simulate the actual vibration behavior of a nuclear fuel rod for springs/dimples in room temperature. When considering the actual contact condition between fuel rod and spring/dimple, if fretting wear progress due to the flow-induced vibration (FIV) under a specific normal load exerted on the fuel rod by the elastic deformation of the spring, the contacting force between the fuel rod and dimple that were located in the opposite side should be decreased. Consequently, the evaluation of developed spacer grids against fretting wear damage should be performed with the results of a cell unit experiments because the contacting force is one of the most important variables that influence to the fretting wear mechanism. Therefore, it is necessary to develop a new type of fretting test rig in order to simulate the actual contact condition. In this paper, the development procedure of a new fretting wear tester and its performance were discussed in detail.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1775-1782
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• 2023

Several reactor physics commissioning experiments were conducted to obtain the neutronic parameters at the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated the calculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuel reactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS first core up to the third core configuration, which has been done experimentally during these commissioning periods. The excess reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to the experiment data. The calculated integral reactivity of the control rod has an error of less than 1.0% compared to the experimental data. The calculated fuel reactivity value is consistent with the measured data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core model is in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2003년도 춘계학술대회논문집
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• pp.82-87
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• 2003

Nuclear fuel with a big slenderness ratio is susceptible to flow-induced vibration under very severe conditions of high temperature, high flow and exposure to irradiation in nuclear reactor. The fuel assembly should, therefore, be designed to escape any resonance due to the vibration during the reactor operation, in particular, in case of the design changes. In addition, the amplitudes due to the grid vibration, the fuel rod vibration and the fuel assembly vibration should be minimized to reduce the grid-to-rod fretting wear. Fuel assembly vibration tests in air at room temperature and in water at high temperature have been performed to investigate fuel vibration behaviors. The frequency and damping during the test in air have been compared to those in water. Through the hydraulic test, the advanced assembly has been evaluated not to be susceptible to any resonance. In addition, the test data from the tests can be used to make fuel model and to evaluate grid-to-rod fretting wear.

• 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.

• Nuclear Engineering and Technology
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• 제28권3호
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• pp.331-338
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• 1996

In the thermal-hydraulic design, the effect of fuel rod bow is quantified tv the rod bow DNBR penalty which is a key design parameter to assure the coolability of fuel assembly in the pressurized water reactor. In this work, a computer program for the evaluation of the rod bow DNBR penalty based on Westinghouse methodology is developed and its application procedure is proposed. The computer simulation is based on the Monte-Carlo method. The qualification of developed computer program is performed by a comparison of calculational result with that given by Westinghouse's document. A new application procedure is built using batch mean and batch standard deviation. The normality of sample population generated by the batch calculation is confirmed by means of a chi-square test for goodness of fit. On the view point of statistics it is effected that the more reliable design value may be produced by the new application procedure.