• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.921-926
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• 1998

CANFLEX 연료봉 다발의 3000 시간 내구성 시험 기간 동안 속도센서를 사용하여 압력관 내부에 장전된 연료봉 다발의 진동을 측정하였다. 압력관 내부에 장전된 연료봉의 진동측정은 고온, 고압, 그리고 공간적 제약 때문에 가속도계나 스트레인 게이지 같은 접촉센서로는 측정할 수 없다. 비접촉 센서를 사용하면 이러한 난점을 해결하고 압력관 내부에 장전된 연료봉 다발의 진동을 측정할 수 있다. 속도센서는 비접촉 센서로서 가우스(gauss)의 크기를 감지하여 전압을 출력하는 센서이지만, 측정거리, 주파수, 그리고 속도와 가우스가 비선형이기 때문에 교정을 한 후에 사용하여야 한다. 본 본문에는 속도센서의 교정방법과 압력관 내부에 고온, 고압의 유체가 흐를 때 발생하는 연료봉 다발의 진동특성을 구하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.306-311
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• 1996

CANFLEX 연료봉 다발을 구성하는 우라늄 펠릿이 장전된 핵연료봉의 공기중 진동특성을 진동 실험과 유한요소 해석을 통하여 구하였다. 유한요소 해석 시에는 우라늄 펠릿의 강성은 무시하고 질량은 지르칼로이 튜브에 부가하며, 연료봉 양단의 용접부위를 단순 지지보로 처리하는 모델을 제시하였다. 이 모델로부터 얻은 해석결과를 진동실험에서 구한 측정값과 비교하였다.

• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.401-409
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• 1991

This work is concerned with the development of an analytical model to predict the friction in longitudinally finned rod bundles. Such bundles are currently considered in KMRR design. The present model assumes the validity of the Law of the Wall over entire flow area. The flow channel area is divided into the interfin region and a number of element channels, and the algebraic form of the Law of the Wall is integrated over each element channel and interfin region to yield an analytic expression for the pressure drop. The model reasonably predicts the 6 fin KMRR data, and overpredicts the 8 fin data about 15 percent.

• Nuclear Engineering and Technology
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• v.23 no.3
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• pp.328-339
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• 1991

A multi-purpose research reactor called KMRR has been developed by Korea Atomic Energy Research Institute(KAERI) to generate a maximum thermal output of 30 MW. As a part of thermal hydraulics study, pressure drop characteristics of the longitudinally finned fuel rod bundles were experimentally investigated in a recirculating water test loop. The present study is focused on the investigation of fin effects on pressure drop and the development of pressure drop correlation for the finned rod bundles in a wide range of flow conditions. Friction factor correlations for each design of the finned rod bundles are developed. The value of friction factor for the finned rod bundles was higher than the analytical solution (64/Re) of laminar circular channel new but became lower than the Blasius equation as Reynolds number was increased.

• Transactions of the Korean Society of Mechanical Engineers
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• v.10 no.1
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• pp.7-14
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• 1986

To predict the fuel clad temperature during the reflooding phase of a LOCA, one may need a knowledge of reflood heat tranfer mechanism in a rod bundle. For this purpose reflooding experiments have been carried out with an electrically heated 3*3 rod bundle. Using the method for the determination of local heat transfer coefficient from the measured wall temperature the parametric effects of coolant flow rate, initial wall temperature, coolant subcooling and heat generation rate on the propagation of rewetting front were investigated. Prediction of the wall temperature histories for these experiments was discussed using REFLUX code with modification of the rewetting temperature correlation. Through this modification, better agreement between experiment and prediction was obtained.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.932-937
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• 1995

1990년 12월 카나다의 Darlington 2호기에서 발생한 핵연료 다발의 양쪽 지지판에 있는 지지 금속판의 파손은 펌프 날개 통과 압력 충격파가 Acoustic 성격으로 중폭되어 연료봉지지판의 파손을 일으킨 것으로 추정되었고 이에 따른 주열수송계통에 대한 ABAQUS를 이용한 Acoustic 해석과 수많은 실험을 거쳐 Acoustic 압력 충격파가 핵연료 다발의 연료봉 지지판 파손 원인임이 입증되었다. 이러한 Acoustic 해석과 실험의 결과로써 Darlington 발전소의 열수송 펌프를 5 날개 펌프에서 7 날개 펌프로 교체시키게 되었으며 그 결과 핵연료 스트링의 축방향 진동을 감소시켜 연료봉 지지판의 파손을 방지하게 되었다. 이러한 사례로 인하여 최근 CANDU형 원자로 열수송 계통의 Acoustic 해석에 대한 연구가 AECL의 Chalk River Laboratory와 COG(CANDU Owners Group)에서 활발하게 진행되고 있다. 이 기고문에서는 매우 새로운 분야로써 현재 이루어지고 있는 CANDU형 원자로 열수송 계통의 Acoustic 해석을 위한 해석 이론과 해석 방법을 간단히 요약 정리하였다.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.97.1-97.1
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• 2005

• The Journal of the Korea Contents Association
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• v.13 no.10
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• pp.46-53
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• 2013

An inspection of nuclear fuel rods should be performed at remoteness from risks of high level radioactivity, and accuracy is required. Currently, inspection of the nuclear fuel rods is operated to monitor the video that recording an original nuclear fuel rods at remoteness because of the risks of radioactivity. In this paper, it is an implementation of the system was carried out in the process according to the image processing inspection of the nuclear fuel rods. The nuclear fuel rods are configured to use a bundle of plurality, in the image processing technology to verify this, the edge detection method is useful. We suggest to DoG technique to add threshold for the nuclear fuel rod edge detections. This is the new technique that optimized DoG. It is to deal with DoG and threshold to dual process. In this way, after detecting an edge of the nuclear fuel rods, by running a nuclear fuel rod inspection algorithm to determine the status of nuclear fuel rods. We applied the system using the new algorithm, and confirmed an excellent characteristic. In this study, it is considered to be able to be carried out more easily and securely inspect of nuclear fuel rods.

• Proceedings of the SAREK Conference
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• 2005.11a
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• pp.300-305
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• 2005

The present experimental study investigates single-phase heat transfer coefficients downstream of support grid in $6{\times}6$ rod bundles. Support grid with split mixing vanes enhance heat transfer in rod bundles by generating it make turbulence. But this turbulence is confined to short distance. Support grid with LSVF mixing vanes enhanced heat transfer to longer distance. The corresponding Reynolds number investigated in the present study is Re=30,000. The heat transfer coefficients are measured using heated and unheated copper sensor.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.149-156
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• 2016

The spent fuel stored in the pool is vulnerable to external impacts, since the severe reactor conditions degrade the structural integrity of the fuel. Therefore an accident during shipping and handling should be considered. In an extreme case, the fuel assembly drop can be happened accidentally during handling the nuclear fuel in the spent fuel pool. The rod failure during such drop accident can be evaluated by calculating the impact force acting on the fuel assembly at the bottom of the spent fuel pool. The impact force can be evaluated with the impact velocity at the bottom of the spent fuel pool. Since fuel rods occupies most of weight and volume of a nuclear fuel assembly, the information of the rods are important to estimate the hydraulic resistance force. In this study, the hydraulic force acting on the $3{\times}3$ short rod bundle model during the drop accident is calculated, and the result is verified by comparing the numerical simulations. The methodology suggested by this study is expected to be useful for evaluating the integrity of the spent fuel.