• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.53-54
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• 2009

The preliminary quantitative analysis of proliferation resistance for the five nuclear fuel cycles demonstrated that the thermal MOX fuel cycle is most vulnerable to proliferation due to the presence of pure $PuO_2$ in the fuel cycle, while the once-through fuel cycle has the highest proliferation resistance. The innovative next generation fuel cycles such as Pyro-SFR and Wet-SFR were found to have similar levels of proliferation resistance to that of the DUPIC fuel cycle which is believed to have proliferation resistance strong enough for commercial deployment. The sensitivity analysis also demonstrated the effectiveness of the proposed methodology in applying to existing and/or newly developing nuclear fuel cycles so as to improve the proliferation resistance characteristic of the fuel cycle systems.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.256-261
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• 2001

Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and to found the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test($400^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test($400^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.781-786
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• 1995

석탄화력발전소 산업부산물인 Fly ash를 이용한 고준위방사성폐기물의 붕규산 유리고화 가능성을 분석하였다. Fly ash SiO$_2$, NaNO$_3$, B$_2$O$_3$에 DUPIC 핵연료 제조공정으로부터 발생되는 모의 scrap waste를 20 wt% 혼합하여, l15$0^{\circ}C$ 에서 3시간 용융시켜 붕규산유리화시켰다. 또한 붕규산유리고화체의 침출성을 평가하기 위하여 2일동안의 soxhlet 침출실험결과 양호한 내침출성을 보였다. 또한 고체폐기물의 안정화물질로 fly ash를 사용할 경우 fly ash 함량을 57%까지 첨가하여도 붕규산유리고화체의 제조가 가능함을 확인하였으며, fly ash의 첨가로 인한 유리화원료 재료비를 30% 까지는 절감시킬 수 있을 것으로 예상된다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11b
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• pp.506-511
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• 1996

Research on the dry etching of UO$_2$ by using fluorine containing gas plasma is carried out for DUPIC (Direct Use of spent PWR fuel In CANDU) process which is taken into consideration for potential future fuel cycle in Korea. CF$_4$/O$_2$ gas mixture is chosen for the reactant gas and the etching rates of UO$_2$ by the gas plasma are investigated as functions of substrate temperature, plasma gas pressure, CF$_4$/O$_2$ ratio, and plasma power, It is tentatively found that the etching rate can reach 1000 monolayers/min. and the optimum CF$_4$/O$_2$ ratio is around 4:1.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.243-251
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• 2015

Nuclear power is currently the second largest power supply method in Korea and the number of nuclear power plants are planned to be increased as well. However, clear management policy for spent fuels generated from nuclear power plants has not yet been established. The back-end fuel cycle, associated with nuclear material flow after nuclear reactors is a collection of technologies designed for the spent fuel management and the spent fuel management policy is closely related with the selection of a nuclear fuel cycle. Cost is an important consideration in selection of a nuclear fuel cycle and should be determined by adding external cost to private cost. Unlike the private cost, which is a direct cost, studies on the external cost are focused on nuclear reactors and not at the nuclear fuel cycle. In this research, external cost indicators applicable to nuclear fuel cycle were derived and quantified. OT (once through), DUPIC (Direct Use of PWR SF in CANDU), PWR-MOX (PWR PUREX reprocessing), and Pyro-SFR (SFR recycling with pyroprocessing) were selected as nuclear fuel cycles which could be considered for estimating external cost in Korea. Energy supply security cost, accident risk cost, and acceptance cost were defined as external cost according to precedent and estimated after analyzing approaches which have been adopted for estimating external costs on nuclear power generation.

• Nuclear Engineering and Technology
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• v.31 no.2
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• pp.133-138
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• 1999

Research on the etching reaction of UO$_2$ with CF$_4$/O$_2$gas mixture plasma is carried out. The reaction rates are investigated as a function of CF$_4$/O$_2$ ratio, plasma power, and substrate temperature. It is found that there exists an optimum CF$_4$/O$_2$ ratio around 4:1 at all temperatures up to 37$0^{\circ}C$ and surface analysis using XPS X-ray Photoelectron Spectroscopy) confirms the result. Peak rate at the optimum gas composition increases with increasing temperature. Highest rate obtained in this study leaches 1050 monolayers/min. at 37$0^{\circ}C$ under r. f. power of 150 W, which is equivalent to about 0.5${\mu}{\textrm}{m}$/min. The rate also increases with increasing r. f. power, thus, higher power and higher substrate temperature will undoubtedly raise the etching reaction rate much further. This reaction seems to be an activated process, whose activation energy will be derived in the following experiments.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.73-73
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• 2009

There are many factors to evaluate nuclear fuel cycle such as safety, public acceptance, economics, etc.. Transparency, proliferation, environment issues, public acceptance and safety are essential to expansion of nuclear industry and proliferation resistance is one of key constraints in the deployment of advanced nuclear energy systems. Proliferation resistance is being considered as one of the most important factors in assessing advanced and innovative nuclear systems. IAEA defmes proliferation resistance as characteristics of nuclear energy system that impedes the diversion or undeclared production of nuclear material [1]. Barriers to proliferation is consist of intrinsic and extrinsic barriers(institutional measures). Intrinsic barriers are characterized in material barriers and technical barriers in general. Material barriers is intrinsic, or inherent, qualities of materials that reduce the inherent desirability or attractiveness of the material as an explosive. Isotopic, chemical, radiological, mass and bulk, detectability barriers are considered as material barriers attributes [2]. Proliferation resistance is examined for several nuclear fuel cycles based on previous study which is focused on the intrinsic barriers [3-4]. Pyroprocessing and DUPIC are considered as reprocessing technologies in Korea and the PWR direct disposal is considered. Comparative assessments of the proliferation attributes and merits of different fuel cycle systems will be performed and the optimal back-end fuel cycle and strategy will be proposed.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.438-446
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• 2003

This paper presents the development of the remotely operated decontamination systems for use in a highly radioactive zone of the DUPIC Fuel Development facility of the Irradiated Material Examination Facility at the Korea Atomic Energy Research Institute. The remotely operated decontamination systems were designed to completely eliminate human interaction with hazardous radioactive contaminants. These decontamination systems are mainly classified into three systems depending on the task environment - a fabrication equipment decontamination system, a hot-cell floor decontamination system, and an isolation room floor decontamination system. A decontamination system for contaminated fabrication equipment utilizes dry ice pellet blasting method to decontaminate contaminated surface of the equipment. The decontamination systems for the hot-cell floor and isolation room floor employ a vacuum cleaning method to decontaminate the contaminated floor and collect loose dry spent nuclear fuel debris and other radioactive waste placed on the floor. The human operator from the out-of-cell performs a series of decontamination tasks remotely by manipulating decontamination systems located in-cell via a handcontroller with the aid of vision feedback information. The environmental, functional and mechanical design considerations, control system and capabilities of the remotely operated decontamination systems at a high radioactive environment are also described.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.53-64
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• 2007

The dynamic release behavior of Cs from high burn-up spent PWR fuel was experimentally performed under the conditions of a thermal treatment process such as voloxidation and sintering conditions. In voloxidation process, influence of the oxidation and reduction atmosphere on the Cs release characteristic using fragment type of spent fuel heated up to $1,500^{\circ}C$ was compared. In sintering process, temperature history effect on Cs release behavior was evaluated using green pellet under 4% $H_2/Ar$ environment. Temperature range for complete Cs release from spent fuel fragment under voloxidation condition was about $800^{\circ}C{\sim}1,200^{\circ}C$, but that of green pellet under the reduction atmosphere was $1,100^{\circ}C{\sim}1,400^{\circ}C$. Key parameters on Cs release behavior from spent fuel was powder formation as well as the diffusion rate of Cs compound to grain boundary and fuel surface.