• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.278-278
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.262.2-262
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• 2002

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.218-227
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• 1986

Studies were conducted to determine the extent of oxidation and same of the mechanical property changes of Zircaloy-4 fuel cladding after it was exposed to hot steam environment. The purpose of these tests was to provide some informations on the embrittlement behavior of CANDU type fuel cladding, which could be experienced under the loss-of-coolant accident conditions. The Zircaloy fuel cladding tubes were exposed in a steam environment at the temperature of 90$0^{\circ}C$, 1,00$0^{\circ}C$. The growth of the ZrO$_2$ layer combined with an oxygen rich $\alpha$-phase layer into the Zircaloy tube material was found as a function of time t and temperature of steam exposure, E=1.1√Dt+0.002 where D is a temperature dependent diffusion coefficient. The tensile strength of the specimens exposed for a short period increased but decreased continuously with further exposure. The circumferential elongation was drastically changed with the exposure time while the hoop strength did't decrease greatly. The X-ray measurement of preferred orientation of the Zircaloy tube material indicated that grains in the as received tube were oriented such that the poles of the basal (0001) planes were predominantly radial, while the poles of the basal plane in the tube materials heattreated at 1,00$0^{\circ}C$ were oriented tangentially. It appears that this reoriented texture may contribute to lessening the decrease of the hoop strength of the heat treated Zircaloy tube material.

• Korean Journal of Materials Research
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• v.4 no.4
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• pp.445-452
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• 1994

Three-axial deformation behavior of the Zircaloy cladding tube under the irradiation condition of the fuel in pressurized water reactor can be analyzed by the anisotropy in the creep and the irra- diation growth, which depends on the texture parameter. A methodology to evaluate the impact of the anisotropic creep and irradiation growth on the strain in each axial direction of the cladding tube has been proposed. Based on the measured strains after irradiation and predicted ones with the help of a fuel performance analysis code, it is found that a tangential strain of the cladding tube is caused mainly by the creep, whereas a axial strain of the cladding is caused mainly by the irradiation growth but with a considerable contribution of the creep at low irradiation.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.204-209
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• 1998

핵연료 펠렛이 장입되어 있는 원전연료봉 피복관은 핵분열성 물질의 외부 유출에 대한 일차 방호벽 역할을 하므로 원전의 안전성을 위해서는 피복관의 구조건전성 확보가 매우 중요하다. 고온, 고압의 운전 조건 속에서 연료봉 피복관은 산화막이 생성 상장하여 연료봉을 취성 파괴시킬 가능성이 있으므로 이를 가동중에 비파괴적으로 측정할 수 있는 방법을 개발할 필요가 있다. 산화막이 존재하는 지르칼로이 피복관에 대한 음파의 공명산란을 이론적으로 모델링하고 수치해석을 수행하였다. 산화막이 피복된 원통형 쉘의 공명산란에서 공명 원주파의 전파 특성은 산화막의 존재 여부와 그 두께 증가에 따라 크게 변화한다. 수치 해석 결과 제 1차 반대칭 (A$_1$) 원주파의 특정 부분파의 경우에는 산화막의 존재에도 불구하고 위상속도가 일정한 특이성을 보였다. 이러한 위상속도 특성을 실험을 통하여 확인하였으며 이 현상을 이용하여 산화막의 두께를 측정할 수 있는 새로운 비파괴 평가 방법을 제안하였다.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.75-81
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• 1990

Experiments to predict the thermal contact conductance between the fuel pellet and cladding have been performed, which is important to determine the temperature distibution within the fuel rod. UO$_2$and Zircaloy-2 are used in these experiments. The measuring apparatus is composed of a presser which controls the contact pressure, a thermometer with 5.5 sheathed thermocouples, a vacuum pump, pellet and cladding rods, and two heating devices, etc. The thermal contact conductances were measured with varying the contact pressure and surface roughnesses of UO$_2$and Zircaloy-2 bars. The results show that an increase in the contact pressure and a decrease of surface roughness resulted in increase of the thermal contact conductance. Finally, a fitting correlation has been established and compared with widely-used correlations.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.475-476
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• 2009

CFD 방법을 이용하여 건식저장 용기내 사용후핵연료 열전달 해석을 수행한 결과 연료봉의 붕괴열에 의한 내부 유체의 자연대류 현상과 상세 핵연료 온도분포를 예측할 수 있음을 확인하였다. 향후에는 다양한 시험조건에서 복사열전달을 포함한 정밀한 CFD 계산을 수행하여 피복관 온도분포의 예측치를 실험결과와 비교할 예정이다.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1999.05a
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• pp.113-118
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• 1999

AMBIDEXTER(Advanced Molten-salt Break-even Inherently-safe Dual-mission Experimental and TEst Reactor)는 고온저압의 Th/$^{233}$ U 불화용융염을 핵연료로 사용하므로 피복관이나 독립된 냉각재 없이 핵연료 자체가 열수송 매체로서 순환하는 원자로시스템개념으로서 저농축 $^{235}$ U 고체 핵연료를 사용하는 기존의 원자력 발전시스템이 안고있는 핵확산과 안전성 등의 고유문제를 해결할 수 있는 혁신형 차세대 원자력 발전시스템이다.(중략)

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.189-196
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• 1988

This study investigates the effect of fuel pellet eccentricity on fuel rod thermal performance under the steady state condition. The governing equations in the fuel pellet and the cladding region are set up in 2-dimensional cylindrical coordinate (r, $\theta$) and are solved by finite element method. The angular-dependent heat transfer coefficient in the gap region is used in order to account for the asymmetry of gap width. Material propeties are used as a function of temperature and volumetric heat generation as a function of radial position. The results show the increase of maximum local heat flux at the cladding outer surface and the decrease of maximum and average fuel temperatures due to eccentricity. The former is expected to affect the uncertainties in the minimum DNBR calculation. The latter two are expected to reduce the possibility of fuel melting and the fuel stored energy. Also, the fuel pellet eccentricity introduces asymmetry in fuel pellet temperature and movement of the location of maximum fuel pellet temperature.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.256-261
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• 2004

To evaluate the effect of spring shape on the fretting wear of nuclear fuel rod, sliding wear tests were performed using three kinds of space grid springs in room temperature air and water. With increasing slip amplitude, wear volume of each spring gradually increased. It is apparently shown that spring with convex shape had a relatively high wear resistance compared with concave shape springs. It is suggested that the ratio of the wear volume to the worn area can be suggested as an efficient and valid parameter to evaluate the wear resistibility of a fuel grid spring.