• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.248-254
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• 2005

This paper presents a quantitative method of sequential separation of $^{90}Sr,\;^{241}Am$ and Pu nuclides with an anion exchange resin and a Sr-Spec resin. The Pu isotopes were purified with an anion exchange resin. The americium and strontium fractions were separated from the matrix elements with an oxalate co-precipitation method. Americium fraction was separated from the strontium fraction with iron co-precipitation method and purified from lanthanides with anion exchange resin. Strontium-90 was purified from other hindrance elements with the Sr-Spec resin after oxalate co-precipitation. The measurement of Pu and Am isotopes was carried out by an ${\alpha}$-spectrometer. Strontium-90 was measured by a liquid scintillation counter. The radiochemical procedure of $^{90}Sr,\;^{241}Am$ and Pu nuclides investigated in this study has been validated by application to IAEA-Reference soils.

• The Journal of Engineering Geology
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• v.3 no.2
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• pp.203-214
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• 1993

Underground Injection Control regulations for Class I injection wells require that the vertical dufusion be considered as a mechanism for transport of contaninants in evaluating containment Due to the low permeability in the confining aquitard, the movement of contaminants over the long term is controlled by the molecular diffusion. The movement can be predicted, using the Fick' S second law of diffusion. The diffusion coefficient in Fick' s law has been determined experimentally in this study. Instantaneous injection of $^{51}Cr$ was used to trace the distribution of Cr(VI) in soil plugs and an analytical solution was applied to calculate the diffusion coefficients. This study shows the effect of environmental factors, such as temperature, chloride concentration, applied amount of $^{51}Cr$, and bulk density of injection formations on diffusion of Cr(VI)

• Nuclear Engineering and Technology
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• v.18 no.4
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• pp.261-266
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• 1986

The purpose of this research was to investigate the de-entrainment of boron for evaporators used in nuclear power plants. The forced circulation and semi-continuous type evaporator was used in the experiment. Cyclone and glass-wool packed column which is supposed to provide good decontamination factor as well as easy maintenance, were selected as de-entrainment devices to be used in the evaporation of radioactive liquid wastees. The de-entrainment device combined with cyclone and glass-wool column has shown overall DF more than 1, 000 for boron.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.97-104
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• 2004

Neutron induced prompt gamma-ray spectroscopy(NIPS) system measures the prompt gamma-ray emitting by the interaction of a neutron with various materials. This system will be of great benefit to scientists worldwide, since it provides the non-destructive measurement of many element in either solid or liquid wastes. In this study, the full-energy-peak (FEP) efficiency calibration for a HPGe detector was constructed in the ${\gamma}$-ray energy range from 80 keV to 8 MeV, using $^{l33}$Ba and >TEX>$^{152}Eu$ RI sources and $ ^{35}Cl(n, ${\gamma}$)^{36}Cl$ thermal neutron captured reaction. The FEP efficiency curve for the higher energies using the $^{35}Cl(n, ${\gamma}$)^{36}Cl$ reaction was normalized with the curve obtained from the RI sources, since the accurate activity of its prompt ${\gamma}$-ray is unknown. The average thermal neutron flux was theoretically calculated using the FEP efficiency curve for the KCl standard solutions. The NIPS system equipped with a ${\gamma}$-${\gamma}$ coincidence setup with two n-type coaxial HPGe detectors was considered in order to reduce the interfering ${\gamma}$-ray background. The FEP efficiency curve for the ${\gamma}$-${\gamma}$ coincidence system was also obtained for full energy range. The performance of the normal and coincidence NIPS system was tested by comparing signal-to-noise ratio in each mode using the reference sample.e.

• Nuclear Engineering and Technology
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• v.6 no.2
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• pp.97-111
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• 1974

The adsorption mechanism of Cs-137 in low level radioactive solution by vermiculite treated with Na ion is studied in order to investigate its effective utilization for the radioactive effluent treatment. The beneficial role of Na-vermiculite is that Na ion can induce the wider c-axis spacing in which Cs ion can be sorbed in vermiculite. Cation exchange capacity and distribution coefficient of cesium seems to be influenced by the variation of c-axis spacing of vermiculite. Comparative identification and detection with the characteristic analyses of X-ray diffraction and electron diffraction patterns, diffrential thermal analysis and electron microscopy of Na-, K- and Cs-vermiculite are studied for the phemomena of Cs adsorption by vermiculite. This importance of the utilization in terms of adsorption and fixation of cesium involving vermiculite is discussed. It is found that the Na-vermiculite is valuable outside charging material for high level radioactive liquid waste storage tank of underground to protect the pollution of the underground water.

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 2000.11a
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• pp.59-59
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• 2000

원자력발전소 1차계통과 격납용기 내부에서 사용되는 주요 부품들은 운전중에 발생한 방사 성 물질들의 침투와 홉착에 의해 오염되어 간다. 이 오염된 부품 및 장비, 공구, 방호복, 방호모자, 작업화 등의 세정과 정비를 위해서는 제염이 선행되어야 한다. 현재의 제염법은 2차 방사성 폐기물을 발생하는 문제점이 있다. 따라서I 2차 폐기물의 발생을 근원적으로 줄일 수 있는 새로운 제염방안이 절실히 요구되고 있는 실정이다. 본 논문에서는 이러한 문제점을 해결할 수 있는 제염법을 개발하기 위해 2가지 방법을 적용하였다. 첫째로, 원자력 발전소에 서 나오는 방사능 오염 세탁물 제염을 위한 액체 및 초임계 이산화탄소를 이용한 방사능 오염물 제염기를 개발하였다. 제염기는 반응기(16 liteer), 회수시스템 그리고 저장용기로 구성되어있다. 세정에 사용된 모든 이산화탄소는 회수되어 재사용 되어지므로 2차 폐기물의 발생을 근원적으로 없앨 수 있다. 제염성능실험결과 제염지수가 목표치보다는 낮았다. 이는 제염 기에 계면활성제와 기계적인 힘을 가한다면 높은 제염지수를 얻을 수 있을 것으로 예상된다. 둘째로, 발전소에서 나오는 오염된 공구나 장비의 세척을 위한 가변형 노즐 드라이 아이스 세척 장치를 개발하였다. 표면세정시 얼음층 형성방지를 위하여 열공급장치를 부착하였다. 유라표면에 지문을 묻혀 실험한 결과 쉽게 제거되었다. 실제 발전소에 있는 P Pump-housing의 표면을 실험한 결과 방사능의 약 40-80%가 제거되었다. 이 장치는 검출기, 제어장치, 용액상에서 세척될 수 없는 장치에 적용할 수 있는 효율적인 세척법이다. 이는 프리프레그의 표면처리 가 충과 충간의 접착강도를 증가시키고 또한 탄소섬유와 에폭시 간의 계면력을 증가시킨데 기인하는 것으로 사려된다.되었으며, duty-on 시간의 증가에 따라 $Cr_2N$ 상의 형성이 점점 많아져 80% duty-on 시간 경우에는 거의 CrN과 $Cr_2N$ 상이 공존하는 것으로 나타났다. 또한 duty-on 시간이 증가할수록 회절피크의 세기가 증가하여 결정화가 더 많이 진행되어짐을 알 수 있었다. 마찬가지로 바이어스 펄스이 주파수에 다른 결정성의 변화도 펄스의 주파수가 증가할수록 박막이 결정성이 좋아지고 $Cr_2N$ 상이 쉽게 형성되었다. 증착 진공도에 따른 결정성은 상대적으로 질소의 농도가 높은 낮은 진공도에서는 CrN 상이 주로 형성되었으며, 반대로 높은 진공도에서는 $Cr_2N$ 상이 많이 만들어졌다. 즉 $1.3{\times}10^{-2}Torr$의 증착 진공도에서는 CrN 상만이 보이는 반면 $9.0{\tiems}1-^{-2}Torr$ 진공도에서부터 $Cr_2N$ 상이 형성되기 시작하여 $5.0{\tiems}10^{-2}Torr$ 진공도에서는 두개의 상이 혼재되어 있음을 알 수 있었다. 박막의 내마모성을 조사한 결과 CrN 박막의 마찰 계수는 초기에 급격하게 증가한 후 0.5에서 0.6 사이의 값으로 큰 변화를 보이지 않았으며, $Cr_2N$ 박막도 비슷한 거동을 보였다.차 이, 목적의 차이, 그리고

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.9-17
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• 2010

The LCC (Liquid Cadmium Cathode) structure to be developed for inhibiting the formation and growth of the uranium dendrite has been known as a key part in the electrowinning process for the simultaneous recovering of uranium and TRU (TRans Uranium) elements from spent fuels. A zinc-gallium (Zn-Ga) experimental system which is able to be functional in aqueous condition and normal temperature has been set up to observe the formation and growth phenomena of the metal dendrites on liquid cathode. The growth of the zinc dendrites on the gallium cathode and the performance of the existing stirrer type and pounder type cathode structure were observed. Although the mechanical strength of the dendrites appeared to be weak in the electrolyte and easily crashed by the various cathode structures, it was difficult to effectively submerge the dendrite into the bottom of the liquid cathode. Based on the results of the aqueous phase experiments, a lab-scale electrowinning experimental apparatus which are applicable to the development of LCC srtucture for the electrowinning process was established and the performance tests of the different types of LCC structure were conducted to prohibit the uranium dendrite growth on LCC surface. The experimental results of the stirrer type LCC structures have shown that they could not effectively remove the uranium dendrites growing at the inner side of the LCC crucible and the performances of the paddle and harrow type LCC structure were similar. Therefore a mesh type LCC structure was developed to push down the uranium dendrites to the bottom of the LCC crucible growing on the LCC surface and at the inner side of the crucible. From the experimental results for the performance test of the mesh type LCC structure, the uranium was recovered over 5 wt% in cadmium without the growth of uranium dendrites. After completion of the experiments, solid precipitates of the bottom of the LCC crucible were identified as an intermetallic compound (UCd11) by the chemical analysis.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.109-114
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• 2009

Absorption spectra for a quantitative analysis of actinide elements such as U(VI) and Pu(V) were measured by using a liquid waveguide capillary cell (LWCC) which has an optical path length of 1.0 meter. In order to investigate radioactive elements, a LWCC is installed in a glove box and is coupled to a spectrophotometer with optical fibers. Limits of detection (LOD) for the system were determined as 0.74 and 0.35 M with molar absorption coefficients of 8.14${\pm}$0.07 (414 nm) and 17.00${\pm}$0.16 (569 nm) $M^{-1}cm^{-1}$ for U(VI) and Pu(V) ions, respectively. The measured LOD values are about 30 times more sensitive when compared to those achievable by using a conventional quartz cell with an optical path length of 1.0 cm. As an application with an enhanced sensitivity, a quantitative analysis for micromolar concentrations of Pu(V) has been performed to decrease the uncertainty in the formation constant of the Pu(VI)-OH complex.

• Korean Journal of Environmental Biology
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• v.34 no.4
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• pp.304-313
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• 2016

One of the issues currently facing nuclear power plants is how to store spent nuclear waste materials which are contaminated with radionuclides such as $^{134}Cs$, $^{135}Cs$, and $^{137}Cs$. Bioremediation processes may offer a potent method of cleaning up radioactive cesium. However, there have only been limited reports on $Cs^+$ tolerant bacteria. In this study, we report the isolation and identification of $Cs^+$ tolerant bacteria in environmental soil and sediment. The resistant $Cs^+$ isolates were screened from enrichment cultures in R2A medium supplemented with 100 mM CsCl for 72 h, followed by microbial community analysis based on sequencing analysis from 16S rRNA gene clone libraries(NCBI's BlastN). The dominant Bacillus anthracis Roh-1 and B. cereus Roh-2 were successfully isolated from the cesium enrichment culture. Importantly, B. cereus Roh-2 is resistant to 30% more $Cs^+$ than is B. anthracis Roh-1 when treated with 50 mM CsCl. Growth experiments clearly demonstrated that the isolate had a higher tolerance to $Cs^+$. In addition, we investigated the adsorption of $0.2mg\;L^{-1}$ $Cs^+$ using B. anthracis Roh-1. The maximum $Cs^+$ biosorption capacity of B. anthracis Roh-1 was $2.01mg\;g^{-1}$ at pH 10. Thus, we show that $Cs^+$ tolerant bacterial isolates could be used for bioremediation of contaminated environments.

• Journal of Radiation Industry
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• v.18 no.2
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• pp.147-153
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• 2024

Radioactive liquid waste generated during operation and overhaul is collected and reused through the radioactive liquid waste treatment system and continuous monitoring system in the nuclear power plant or discharged to the outside if it satisfies the limit within the control and monitoring. However, there are concerns about boric acid management, which controls the power output of nuclear power plants in radioactive liquid waste. Due to the behavior of boric acid, it is difficult to remove it in the existing liquid radwaste system, and the concentration of boric acid water discharged tends to be higher than the natural state of 5 ppm, so additional facilities should be considered. Therefore, this study aims to evaluate the radiological effects of radioactive waste treatment facilities that are under development and use them as a basis for managing worker exposure and evaluating the safety of facilities in the future.