• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.277-277
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• 1999

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.05a
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• pp.840-843
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• 2002

The dry pulverizing/Mixing device is used to deal with the spent fuels for the safe disposal. The separated pellets from hulls by a slitting device are put and oxidized from UO$_2$ solid pellet to U$_3$O$\_$8/ powder in the device. The device have been developed based on a voloxidation method which is one of several dry de-cladding methods. We have benchmarked dry de-cladding methods, analyzed applicability to the advanced spent fuel management process, integrated and compared several configuration, and finally derived detailed specifications proper to requirements for the device. Also, thermal characteristics of the device such as thermal stress and strain have been analyzed by the commercial software, 1-DEAS, and the reliability of the results have been verified by the KOLAS(Korea Laboratory Accreditation Scheme). The UO$_2$ solid pellets are put in the device which has a capacity of 20 kgHM per a batch, heated up about 600$^{\circ}C$ in the air environment. Then, the UO$_2$ solid pellets are oxidized into the U$_3$O$\_$8/ powder, and the powder is collected in a special vessel. The device has been designed and developed as fellows: the multi-staged fine hole meshes are used to reduce the size of the powder gradually, heat and air(oxygen) are supplied continuously to reduce the reaction time, and slight vibration effect are applied to collect powder cling to the device.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.820-828
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• 2018

The growing nuclear threat has amplified the need for developing diverse and accurate nuclear forensics analysis techniques to strengthen nuclear security measures. The work presented here is part of a research effort focused on developing a methodology for reactor-type discrimination of weapons-grade plutonium. To verify the developed methodology, natural $UO_2$ fuel samples were irradiated in a thermal neutron spectrum at the University of Missouri Research Reactor (MURR) and produced approximately $20{\mu}g$ of weapons-grade plutonium test material. Radiation transport simulations of common thermal reactor types that can produce weapons-grade plutonium were performed, and the results are presented here. These simulations were needed to verify whether the plutonium produced in the natural $UO_2$ fuel samples during the experimental irradiation at MURR was a suitable representative to plutonium produced in common thermal reactor types. Also presented are comparisons of fission product and plutonium concentrations obtained from computational simulations of the experimental irradiation at MURR to the nondestructive and destructive measurements of the irradiated natural $UO_2$ fuel samples. Gamma spectroscopy measurements of radioactive fission products were mostly within 10%, mass spectroscopy measurements of the total plutonium mass were within 4%, and mass spectroscopy measurements of stable fission products were mostly within 5%.

• Journal of the Korean Society for Precision Engineering
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• v.30 no.1
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• pp.113-119
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• 2013

To develop a new nuclear fuel, it needs to make a test fuel rod and carry out burn-up test in the test loop of a research reactor to check the irradiation characteristics of the nuclear fuel. At that time, several sensors such as thermocouples, LVDTs and SPNDs are needed to be attached in and out of the fuel rod and connect them with instrumentation cables. Then, the instrumentation cables deliver the signals measured by the sensors to the measuring device located outside of the reactor pool. In particular, to install a thermocouple in a fuel rod, it needs to drill off holes on the alumina blocks and sintered $UO_2$ pellets. However, because the hardness of a sintered $UO_2$ pellet is 700 Hv (or HRC 61) and that of an alumina block is 1480 Hv, a special drilling machine which adapts a diamond coated drill bit had developed. In this study, several case experiments have been carried out to find an optimal drilling condition of the drilling machine. And, using the optimal drilling condition, minimum numbers of the holes that a drill bit can drill off are verified.

• Nuclear Engineering and Technology
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• v.35 no.1
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• pp.14-24
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• 2003

This work investigates the effects of Cr$_2$O$_3$ and oxygen potential on grain growth and densification of UO$_2$ pellets. Powder mixtures of UO$_2$ and 0.03-0.4wt% Cr$_2$O$_3$ were pressed and sintered in 3 different gas atmospheres: the $H_2O$-to-H$_2$ ratios were 5$\times$10$^{-4}$ , 1$\times$10$^{-2}$ and 3$\times$10$^{-2}$ In the first gas atmosphere the Cr$_2$O$_3$ contents below 0.2 wt% have an insignificant effect on grain size, but the Cr$_2$O$_3$ contents more than 0.3 wt% promote grain growth in the inner zone of a pellet but not in the outer zone. In both the second and third atmospheres, the grain size increases with the Cr$_2$O$_3$ content. With the same level of Cr$_2$O$_3$ content the grain size is larger in the second atmosphere than in the third. Sintering behavior and developed microstructure are discussed in terms of the reduction of C$r^2$O$^3$ to Cr, the dissolution of C$r^2$O$^3$ in UO$_2$, and liquid phase sintering.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.335-343
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• 1999

The effects of N $b_2$ $O_{5}$ and oxygen potential on the densification and grain growth of U $O_2$ fuel have been investigated.0.3 wt% N $b_2$ $O_{5}$ -doped U $O_2$fuel pellets were sintered at 1$700^{\circ}C$ for 4 hours in sintering atmospheres which have various ratios of $H_2O$ to $H_2$ gas. Compared with those of undoped U $O_2$ pellets, the sintered density and grain size of the 0.3 wt% N $b_2$ $O_{5}$ -doped U $O_2$ pellet increase under the $H_2O$/ $H_2$ gas ratio of 5.0$\times$10$^{-3}$ to 1.0$\times$10$^{-2}$ and under the $H_2O$/ $H_2$gas ratio of 5.0$\times$10$^{-3}$ to $1.5\times$10$^{-2}$ , respectively. The sintering of U $O_2$fuel pellets containing 0.1 wt% to 0.5 wt% N $b_2$ $O_{5}$ was carried out at 168$0^{\circ}C$ for 4 hours. The enhancing effect of N $b_2$ $O_{5}$ on the sintered density and grain size becomes larger as the N $b_2$ $O_{5}$ content increases. The solubility limit of N $b_2$ $O_{5}$ in U $O_{2}$ seems to be between 0.3 wt% and 0.5 wt%, and beyond the solubility limit the second phase whose composition corresponds near to N $b_2$U $O_{6}$ is precipitated on grain boundary. The enhancement of densification and grain growth in U $O_2$ is attributed to the increased concentration of a uranium vacancy which is formed by the interstitial N $b^{4+}$ ion in the U $O_2$ lattice.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.101-116
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• 2017

The interaction of U(VI) (hexavalent uranium) species with natural organic matter (NOM) in KURT (KAERI Underground Research Tunnel) groundwater is investigated using a laser spectroscopic technique. The luminescence spectra of the NOM are observed in the ultraviolet and blue wavelength regions by irradiating a laser beam at 266 nm in groundwater. The luminescence spectra of U(VI) species in groundwater containing uranium concentrations of $0.034-0.788mg{\cdot}L^{-1}$ are measured in the green-colored wavelength region. The luminescence characteristics (peak wavelengths and lifetime) of U(VI) in the groundwater agree well with those of $Ca_2UO_2(CO_3)_3(aq)$ in a standard solution prepared in a laboratory. The luminescence intensities of U(VI) in the groundwater are weaker than those of $Ca_2UO_2(CO_3)_3(aq)$ in the standard solution at the same uranium concentrations. The luminescence intensities of $Ca_2UO_2(CO_3)_3(aq)$ in the standard solution mixed with the groundwater are also weaker than those of $Ca_2UO_2(CO_3)_3(aq)$ in the standard solution at the same uranium concentrations. These results can be ascribed to calcium-U(VI)-carbonate species interacting with NOM and forming non-radiative U(VI) complexes in groundwater.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.05a
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• pp.151-152
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• 2007

ACP(Advanced Spent Fuel Conditioning Process)의 금속전환로에 $U_{3}O_{8}$을 공급하기 위하여 20 kgHM/batch의 $UO_{2}$ 펠릿(pellets)을 처리할 수 있는 공기산화로가 개발되고 있다. 그림 1은 산소농도 조절이 가능한 공기산화로이다. 공기산화로 이전의 공정인 슬리팅 장치에서 탈피복된 $UO_{2}$ 펠릿은 공기산화로로 운반되고, $500^{\circ}C$온도에서 공기를 공급하여 일정한 입도범위의 균질한 $U_{3}O_{8}$을 만든다. 그리고 다음공정의 금속전환장치로 이동된다. 본 논문에서는 모의연료의 산화에 대한 정확한 산소농도를 측정하고자 한다. 이를 위해서 갈바닉 센서와 지르코니움 센서가 사용되었고, 그 특성이 비교되었다. 14종의 금속 산화물이 혼합된 모의연료를 제조하여 산화실험이 수행되었으며, 시간변화에 따라 산소농도가 측정되었다. 산소농도 컨트롤러와 산소 센서를 사용한 공기산화로는 산소조절기에 의해 산소농도 100%까지 측정될 수 있다. 그림 2는 공기산화로의 산소농도를 조절할 수 있는 산소농도 측정시스템이다. 유량조절기(Mass Flow Controller)를 사용하여 질소와 산소의 혼합비를 변화시킬 수 있다. 또한 산소농도 측정시스템은 측정된 산소농도 값을 이용하여 $UO_{2}$의 산화시간을 계산하기 위하여 제작하였다. 산화시간 계산방법은 다음과 같다. 산소와 질소의 가스는 각각 40 L의 압력 봄베에 의해서 산소농도를 조절할 수 있는 공기산화로의 산소농도 측정시스템 안으로 유입된다. 유입된 산소와 질소의 배합은 컨트롤시스템 안에 있는 산소 유량조절기와 질소 유량 조절기를 사용하여 조절하며, 일정하게 혼합된 산소농도는 장치의 입구와 출구에서 산소 센서에 의해서 측정된다. 투입된 $UO_{2}$ 펠릿이 $500^{\circ}C$에서 반응하면서 공기산화로의 내부에 있는 산소농도가 감소된다. 이때 초기에 같았던 입력과 출력 농도가 시간의 흐름에 따라 감소되며, 펠릿이 완전히 산화됨과 동시에 출력 산소농도가 입력농도와 다시 같아질 때까지 소요된 구간이 산화시간이 된다.

• Journal of the Korean Chemical Society
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• v.37 no.9
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• pp.813-819
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• 1993

An ICP-AES system directly connected with a separation column was used in order to determine the trace elements in $UO_2$ powder promptly and reduce the volume of the waste solution. The outlet of a separation column, which was filled with Teflon powder ($330\;{\mu}m$) coated with tri-n-butyl phosphate (TBP) as extractant, was directly connected with sample injection tube of ICP-AES. Eleven elements including molybdenum in $UO_2$ powder were separated and determined simultaneously. Recoveries of these elements were $91{\sim}110%$ and these results were agreed with those of solvent extraction methods. This method was applicable to quality control in manufacturing nuclear fuel.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1035-1036
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• 2004