• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.329-334
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• 1996

손상 핵연료의 온도 및 산소대 금속 비율의 변화모형을 연구하였다. 열역학적 분석과 산화과정에 대한 분석을 통해 손상핵연료에서의 핵연료 온도와 핵연료내 O/U값 변화를 기술함으로써 결함발생에 의한 핵연료내 냉각수 침투는 Gap conductance를 떨어뜨리고 소결체 산화에 따른 O/U값 증가로 열전달특성의 저하를 가져와 핵연료의 온도를 상승시킨다는 결과를 얻어냈다.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.811-824
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• 1995

The CANDU element bowing is attributed to actions of both the thermally induced bending moments and the bending moment due to hydraulic drag and mechanical loads, where the bowing is defined as the lateral deflection of an element from the axial centerline. This paper consider only the thermally-induced bending moments which are generated both within the sheath and the fuel and sheath by an asymmetric temperature distribution with respect to the axis of an element The generalized and explicit analytical formula for the thermally-induced bending is presented in con-sideration of 1) bending of an empty tube treated by neglecting the fuel/sheath mechanical interaction and 2) fuel/sheath interaction due to the pellet and sheath temperature variations, where in each case the temperature asymmetries in sheath are modelled to be caused by the combined effects of (i) non-uniform coolant temperature due to imperfect coolant mixing, (ii) variable sheath/coolant heat transfer coefficient, (iii) asymmetric heat generation due to neutron flux gradients across an element and so as to inclusively cover the uniform temperature distributions within the fuel and sheath with respect to the axial centerline. As the results of the sensitivity calculations of the element bowing with the variations of the parameters in the formula, it is found that the element bowing is greatly affected relatively with the variations or changes of element length, sheath inside diameter, average coolant temperature and its variation factor, pellet/sheath mechanical interaction factor, neutron flux depression factor, pellet thermal expansion coefficient, pellet/sheath heat transfer coefficient in comparison with those of other parameters such as sheath thickness, film heat transfer coefficient, sheath thermal expansion coefficient and sheath and pellet thermal conductivities.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.207-212
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• 1996

DUPIC 핵연료 제조를 위해서 PWR 사용후핵연료의 피복관과 소결체를 분리하는 방법이 검토되었다. 약 50 cm의 PWR 사용후핵연료의 길이방향에 일정한 간격으로 구멍을 뚫어 산화하는 방법, 10∼20 m의 길이로 핵연료봉을 절단하여 산화하는 방법, 그리고, 핵연료봉의 길이 방향에서 피복관을 slitting하여 산화하는 방법에 대해 실험을 수행하였다. 실험 결과들로 보아 DUPTC 핵연료 제조 공정에 가장 적합한 방법으로는 핵연료봉 길이방향으로 slitting하여 산화하는 것이 가장 타당한 것으로 판명되었다.

• Journal of Energy Engineering
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• v.10 no.3
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• pp.278-285
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• 2001

In this study, the effects of the variables on sintering of simulated fuel to simulate the spent fuel are described. Mainly, the effects of compaction pressure, sintering temperature and time on the density of pellet are described. The experimental is performed with compaction pressure of 1 ton/$\textrm{cm}^2$~4 ton/$\textrm{cm}^2$, sintering temperature of 167$0^{\circ}C$, 173$0^{\circ}C$ and 178$0^{\circ}C$ and sintering time of 4 hr, 8 hr and 24 hr. The green density of simulated fuel is proportional to the one third power of compaction pressure and the sintered density is 90.5~99.6% of theoretical density. The grain growth exponent and activation energy of simulated fuel is 2.5 and 287.97 kJ/mol, respectively.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.90-99
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• 1994

One of the major phenomena occurring in defective fuel rods is the oxidation of UO$_2$ fuel pellets from UO$_2$ to UO$_{2+}$x/ by the oxygen Produced from the dissociation of the steam in the Pellet-to-clad gap, which leads to the enhancement of fission gas release. In this paper, the oxidation kinetics of defective fuel rods was analyzed on the basis of operating conditions of the reactor and defective fuel rod itself. Oxidation kinetics of the fuel pellet was determined under the assumption that the gap is filled with the saturated steam of 150 atm and an enhancement factor for fission gas release was introduced to take into account the effect of fuel oxidation on fission gas release. Comparison with experimental data shows that the enhancement factor predicts well the increased fission gas release due to the oxidation of UO$_2$fuel pellets.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.424-429
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• 1993

Numeric characters are printed at the end part of nuclear fuel rod containing nuclear pellets. Fuel rods are discriminated and managed systematically by these characters in the process of producing fuel assembly. The characters are also used to examine manufacturing process of fuel rods in the survey of burnup efficiency as well as in inspection of irradiated fuel rod. Therefore automatic character recognition is one of the most important technologies in automatic manufacture of fuel assembly. In this study, character recognition system is developed. In the developed system, mesh feature extracted from each character written in the fuel rod has been compared with reference feature value stored in database, and the character is thus identified. In the result of experiment, 95.83 percent recognition rate is achievable.

• Resources Recycling
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• v.11 no.4
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• pp.3-10
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• 2002

The characteristics of dry and wet milled powder prepared by 1 cycle OREOX (oxidation and reduction of oxide fuels) treatment were investigated using the simulated spent fuel pellet. Sintered pellets simulating spent nuclear fuel burned in reactor were fabricated from $UO_2$ powder using as a starting material in fabrication of nuclear fuel. The 1 cycle OREOX-treated powder was prepared by only one path of oxidation md reduction of the simulated pellet. Powder having average particle size of less than 1 $\mu\textrm{m}$ could be easily obtained by dry milling, but not be achieved by wet milling. And, specific surface area of dry milled pow-der was higher than that of wet milled powder. Dry milled powder formed loose agglomerate, while wet milled powder showed the shape of irregular and angular particles. Dry milled powder provided higher green density, resulting in higher sintered density of higher than 95% TD and average grain size of larger than 8 $\mu\textrm{m}$ satisfying the standard specification of sintered pellets.

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2005.11a
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• pp.25-26
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• 2005

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.10a
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• pp.161-162
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• 2012

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.300.1-300
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• 2001