• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.1-7
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• 2006

The spectroscopic characteristics of Pu (III, IV, V, VI) in the HCl media were investigated by measuring Pu oxidation states using a UV-Vis-NIR spectrophotometer (400-1200 nm) after adjusting Pu oxidation states with oxidation/reduction reagents. Pu in stock solution was reduced to Pu(III) with $NH_2OH$ HCl, and oxidized to Pu(IV) and Pu(VI) with $NaNO_2$ and $HClO_4$, respectively. Also, Pu(V) was adjusted in the Pu(VI) solution with $NH_2OH$ HCl. The major absorption peaks of Pu (IV) and Pu(III) were measured in the 470 m and 600 nm, respectively. The major absorption peaks of Pu (VI) and Pu(V) were measured in the 830 nm and 1135 nm, respectively. There was not found to be significant changes of UV-Vis absorption spectra for Pu(III), Pu(IV) and Pu(VI) with aging time, except that an unstable Pu(V) immediately reduced to Pu(III).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.245-261
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• 2019

Radioactive contaminant from a nuclear facility moves to the ecosystem by run-off or groundwater flow. Among the two mechanisms, contaminant plume through a river can be easily detected through a surface water monitoring system, but radioactive contaminant transport in groundwater is difficult to monitor because of lack of information on flow path. To understand the contaminant flow in groundwater, understanding of the geo-environment is needed. We suggest a method to decide on monitoring location and points around an imaginary nuclear facility by using the results of site characterization in the study area. To decide the location of a monitoring well, groundwater flow modeling around the study area was conducted. The results show that, taking account of groundwater flow direction, the monitoring well should be located at the downstream area. Also, monitoring sections in the monitoring well were selected, points at which groundwater moves fast through the flow path. The method suggested in the study will be widely used to detect potential groundwater contamination in the field of oil storage caverns, pollution by agricultural use, as well as nuclear use facilities including nuclear power plants.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.329-338
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• 2019

Suction of unsaturated bentonite buffers is a very important input parameter for hydro-mechanical performance assessment and design of an engineered barrier system. This study analyzed suction measurement techniques and constitutive models of unsaturated porous media reported in the literature, and suggested suction measurement techniques and constitutive models suitable for bentonite buffer in an HLW repository. The literature review showed the suction of bentonite buffer to be much higher than that of soil, as measured by total suction including matric suction and osmotic suction. The measurement methods (RH-Cell, RH-Cell/Sensor) using a relative humidity sensor were suitable for suction measurement of the bentonite buffer; the RH-Cell /Sensor method was more preferred in consideration of the temperature change due to radioactive decay heat and measurement time. Various water retention models of bentonite buffers have been proposed through experiments, but the van Genuchten model is mainly used as a constitutive model of hydro-mechanical performance assessment of unsaturated buffers. The water characteristic curve of bentonite buffers showed different tendencies according to bentonite type, dry density, temperature, salinity, sample state and hysteresis. Selection of water retention models and determination of model input parameters should consider the effects of these controlling factors so as to improve overall reliability.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.1-11
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• 2004

The hot cell facility for research activities related to the lithium reduction of spent fuel, which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations peformed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}, 2.97{\times}10^{-3} and 1.01{\times}10{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}, 2.99{\times}10^{-3} and 7.88{\times}10^{-2}$ mSv/h, respectively, The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01 mSv/h for the operation area and 0.15 mSv/h for the service (maintenance) area.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.161-170
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• 2006

PFC(Perfluorocarbon) decontamination process is one of best methods to remove hot particulate adhered on the inner surface of hot cell and surface of equipment in hot cell. It was necessary to develop a filtration equipment to reuse the PFC waste solution generated on PFC decontamination due to the high cost of PFC solution and for minimization of the volume of second waste solution. The filtration equipment was developed to remove hot particulate in PFC waste solution. It was made suitable size and weight in consideration of hot cell gate and crane. And it has wheels for easy movement. Flux of the filtration equipment decreased with particulate concentration increase. It consists of pre-filter($1.4{\mu}m$) and final-filter($0.2{\mu}m$) for protection of the flux decrease along filtration time. It treatment capacity of waste solution is 0.2 L/min.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.1-11
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• 2012

The equation of the step-drawdown test "$s_w=BQ+CQ^p$" written by Rorabaugh (1953) is suitable for drawdown increased non-linearly in the fractured rocks. It was found that value of root mean square error (RMSE) between observed and calculated drawdowns was very low. The calculated $C$ (well head loss coefficient) and $P$ (well head loss exponent) value of well head losses ($CQ^p$) ranged $3.689{\times}10^{-19}{\sim}5.825{\times}10^{-7}$ and 3.459~8.290, respectively. It appeared that the deeper depth in pumping well the larger drawdowns due to pumping rate increase. The well head loss in the fractured rocks, unlike that in porous media, is affected by properties of fractures (fractures of aperture, spacing, and connection) around pumping well. The $C$ and $P$ value in the well head loss is very important to interpret turbulence interval and properties of high or low permeability of fractured rock. As a result, regression analysis of $C$ and $P$ value in the well head losses identified the relationship of turbulence interval and hydraulic properties. The relationship between $C$ and $P$ value turned out very useful to interpret hydraulic properties of the fractured rocks.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.95-102
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• 2006

Alpha spectrometry was studied for the determination of $^{237}Np$ in spent nuclear fuel samples. The optimum condition for the electrodeposition of $^{237}Np$ was obtained as follows : for $1{\sim}1.5$ hour of deposition time, at the current intensity of $1.2{\sim}1.5$ A and at sodium sulfate electrolyte without organic additive. The deposition yield and its reproducibility on $^{237}Np$ was decreased as the amount of $^{237}Np$ decreased from 4.16 Bq down to 0.0264 Bq(1ng). The recovery yield of $^{237}Np$ determined by alpha spectrometry after separation in synthetic solution was $98.8{\pm}5.1%$(n=4). The contents of $^{237}Np$ in spent nuclear fuel samples were determined and the result showed an agreement within 10% of a difference between the measurement and the calculation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.3
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• pp.235-242
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• 2015

A commercially available copper mesh was investigated as an iodine off-gas capturing medium for pyroprocessing, with an aim to replace costly silver based adsorbents. Theoretical calculation results suggested that the reaction between metallic copper and gaseous iodine will occur spontaneously to produce copper iodide in the temperature range of 100 ~ 500℃. The effect of the reaction temperature on iodine capturing efficiency was investigated by experimentation, and it was found that 5 and 6 wt% of iodine (initial mass 2.0 g) was captured by a single copper mesh (0.26 g) at 300 and 400℃, respectively. The repeated experimental results also suggested that copper utilization can be increased with the help of the spontaneous detachment of the reaction product (CuI) from a copper mesh. The formation of the CuI phase was confirmed using the X-ray diffraction technique, and the surface morphology of the reaction product was observed using scanning electron microscopy.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.11-23
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• 2003

A study on the electrolytic dissolution of SUS-304 and Inconel-600 specimen was carried out in neutral salt electrolyte to evaluate the applicability of electrochemical decontamination process for recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant in Korea. Although the best electrolytic dissolution performance for the specimens was observed in a Na2s04 electrolyte, a NaNO$_3$ neutral salt electrolyte, in which about 30% for SUS-304 and the same for Inconel-600 in the weight loss was shown in comparison with that in a Na$_2$SO$_4$ solution, was selected as an electrolyte for the electrochemical decontamination of metallic wastes with the consideration on the surface of system components contacted with nitric acid and the compatibility with lagoon wastes generated during the facility operation. The effects of current density, electrolytic dissolution time, and concentration of NaNO$_3$ on the electrolytic dissolution of the specimens were investigated. On the basis of the results obtained through the basic inactive experiments, electrochemical decontamination tests using the specimens contaminated with uranium compounds such as UO$_2$, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion facility were performed in 1M NaNO$_3$ solution with the current density or In mA/$\textrm{cm}^2$. it was verified that the electrochemical decontamination of the metallic wastes contaminated uranium compounds was quite successful in a NaNO$_3$ neutral salt electrolyte by reducing $\alpha$ and $\beta$ radioactivities below the level of self disposal within 10 minutes regardless of the type of contaminants and the degree of contamination.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.171-177
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• 2007

This study suggested a new method to stabilize molten salt wastes generated from the pyre-process for the spent fuel treatment. Using conventional sol-gel process, $SiO_2-Al_2O_3-P_2O_5$ (SAP) inorganic material that is reactive to metal chlorides were prepared. In this paper, the reactivity of SAP with the metal chlorides at $650{\sim}850$, the thermal stability of reaction products and their leach-resistance under the PCT-A test method were investigated. Alkali metal chlorides were converted into metal aluminosilicate($LixAlxSi1-_xO_{2-x}$) and metal phosphate($Li_3PO_4\;and\;Cs_2AlP_3O_{10}$) While alkali earth and rare earth chlorides were changed into only metal phosphates ($Sr_5(PO_4)_3Cl\;and\;CePO_4$). The conversion rate was about $96{\sim}99%$ at a salt waste/SAP weight ratio of 0.5 and a weight loss up to $1100^{\circ}C$ measured by thermogravimetric analysis were below 1wt%. The leach rates of Cs and Sr under the PCT-A test condition were about $10^{-2}g/m^2\;day\;and\;10^{-4}g/m^2\;day$. From these results, it could be concluded that SAP can be considered as an effective stabilizer for metal chlorides and the method using SAP will give a chance to reduce the volume of salt wasteform for the final disposal through further researches.