• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.271-280
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• 2013

Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. And concrete waste was generated about 800 drums of 200 L. The conditioning of concrete waste is needed for final disposal. The concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled void space after concrete rubble pre-placement into 200 L drum. Thus, this research has developed an optimizing mixing ratio of concrete waste, water, and cement and has evaluated characteristics of a cement waste form to meet the requirements specified in disposal site specific waste acceptance criteria. The results obtained from compressive strength test, leaching test, thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10wt% as the optimized mixing ratio. Also, the compressive strength of cement waste form was satisfied that including fine powder up to maximum 40wt% in concrete debris wastes about 75%. As a result of scale-up test, the mixture of concrete waste, water, and cement is 75:10:15wt% meet the satisfied compressive strength because the free water increased with and increased in particle size.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.259-269
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• 2013

Applicability of ammonium molybdophosphate/iron oxides-polyacrylonitrile (AMP/IO-PAN) composites on the removal of radionuclides in the radioactive wastewater generated from nuclear power plants was investigated. The composites were characterized using the following analytical techniques: X-ray diffraction (XRD), Fourior transform-infrared (FT-IR) spectroscopy, scanning electron microscopy (SEM), particle size analyzer (PSA), nitrogen adsorption-desorption and magnetic property measurement system (MPMS). 10wt% of AMP/IO-PAN composite has a saturation magnetization of 2.038 emu/g. Single-solute sorptions of Co, Sr and Cs onto 10wt% of AMP/IO-PAN composite were investigated. The maximum sorption capacities ($Q^0$) predicted by the Langmuir model on 10wt% of AMP/IO-PAN composite were 0.097, 0.086 and 0.66 mmol/g for Co, Sr and Cs, respectively. The maximum sorption capacities ($Q^0$) of Cs predicted by Langmuir model on 0, 10, 20 and 30wt% of AMP/IO-PAN composites were 0.702, 0.655, 0.602 and 0.559 mmol/g, respectively. The maximum sorption capacities ($Q^0$) of Cs decreased with increasing the iron oxide content in the AMP/IO-PAN composites.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.3
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• pp.171-180
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• 2015

This study was performed on the removal of Cs, one of the main high- radioactive nuclides contained in the high-radioactive seawater waste (HSW), by adsorption with IE911 (crystalline silicotitanate type). For the effective removal of Cs and the minimization of secondary solid waste generation, adsorption of Cs by IE911 (hereafter denoted as IE911-Cs) was effective to carry out in the m/V (ratio of absorbent weight to solution volume) ratio of 2.5 g/L, and the adsorption time of 1 hour. In these conditions, Cs and Sr were adsorbed about 99% and less than 5%, respectively. IE911-Cs could be also expressed as a Langmuir isotherm and a pseudo-second order rate equation. The adsorption rate constants (k₂) were decreased with increasing initial Cs concentrations and particle sizes, and increased with increasing ratios of m/V, solution temperatures and agitation speeds. The activation energy of IE911-Cs was about 79.9 kJ/mol. It was suggested that IE911-Cs was dominated by a chemical adsorption having a strong bonding form. From the negative values of Gibbs free energy and enthalpy, it was indicated that the reaction of IE911-Cs was a forward, exothermic and relatively active at lower temperatures. Additionally, the negative entropy values were seen that the adsorbed Cs was evenly distributed on the IE911.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.317-325
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• 2020

The Korea Radioactive Waste Agency plans to expand the storage capacity of radioactive waste by constructing a radioactive waste inspecting building to solve the problem of the lack of inspection space and drum-handling space in the radioactive waste receipt and storage building for the first-stage disposal facility. In this study, the exposure doses of radiation workers that handle new disposal containers for decommissioning waste in the storage areas of the radioactive waste inspecting building were calculated using the Monte Carlo N-particle transport code. The annual collective dose was calculated as a total of 84.8 man-mSv for 304 new disposal containers and an estimated annual 306 working hours for the radiation work. When the 304 new disposal containers (small/medium type) were stored in the storage areas, it was found that 25 radiation workers should be involved in acceptance/disposal inspection, and the estimated exposure dose per worker was calculated as an average annual value of 3.39 mSv. When the radiation workers handle the small containers in high-radiation dose areas, the small containers should be shielded further by increasing the concrete liner thickness to improve the work efficiency and radiation safety of the radiation workers. The results of this study will be useful in establishing the optimal radiation working conditions for radiation workers using the source term and characteristics of decommissioning waste based on actual measurements.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.1-14
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• 2017

For the removal of cesium (Cs) from high radioactive/high salt-laden liquid waste, this study synthesized a highly efficient composite adsorbent (potassium cobalt ferrocyanide (PCFC)-loaded chabazite (CHA)) and evaluated its applicability. The composite adsorbent used CHA, which could accommodate Cs as well as other molecules, as a supporting material and was synthesized by immobilizing the PCFC in the pores of CHA through stepwise impregnation/precipitation with $CoCl_2$ and $K_4Fe(CN)_6$ solutions. When CHA, with average particle size of more than $10{\mu}m$, is used in synthesizing the composite adsorbent, the PCFC particles were immobilized in a stable form. Also, the physical stability of the composite adsorbent was improved by optimizing the washing methodology to increase the purity of the composite adsorbent during the synthesis. The composite adsorbent obtained from the optimal synthesis showed a high adsorption rate of Cs in both fresh water (salt-free condition) and seawater (high-salt condition), and had a relatively high value of distribution coefficient (larger than $10^4mL{\cdot}g^{-1}$) regardless of the salt concentration. Therefore, the composite adsorbent synthesized in this study is an optimized material considering both the high selectivity of PCFC on Cs and the physical stability of CHA. It is proved that this composite adsorbent can remove rapidly Cs contained in high radioactive/high salt-laden liquid waste with high efficiency.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.53-59
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• 2004

A variety of factors such as the pattern of intake (acute or chronic), monitoring interval and the characteristics of the radionuclides could have a significant influence on the estimates for the intake and internal dose. The relative differences of the assessed intakes based on the assumption of an acute intake to that of a chronic intake were evaluated by using the predicted bioassay quantity in the whole body or organs for an acute and chronic intake through the inhalation of $^{125}$ I, $^{137}$ C, $^{235}$ U with the AMAD of 1 ${\mu}{\textrm}{m}$ and 5 ${\mu}{\textrm}{m}$ for the monitoring intervals of 7, 14, 30, 60, 90, 120, 180, 360 days, respectively, The relative difference of the assessed intakes based on the intake pattern is affected by the monitoring interval, radionuclide and absorption type, but the particle size has little influence on the difference of the assessed intakes based on the intake pattern. The maximum monitoring interval, which is defined as the monitoring interval that the relative difference of the assessed intakes based on the assumption of an acute intake to that of a chronic intake is less than 10%, is 60 d for $^{125}$ I with Type F, 180 d for $^{137}$ C with Type F, 90 d for $^{235}$ U with Type M, and 360 d for $^{235}$ U with Type S. It was concluded that an intake pattern has little influence on the estimates of the assessed intake in the case where the monitoring interval is shorter than the maximum monitoring interval for each radionuclide.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.295-303
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• 2012

Based on the data observed and analyzed on a groundwater flow system in the KURT (KAERI Underground Research Tunnel) site, the transport of radionuclides, which were assumed to be released at the supposed position, was calculated on the time-domain. A groundwater pathway from the release position to the surface was identified by simulating the groundwater flow model with the hydrogeological characteristics measured from the field tests in the KURT site. The elapsed time when the radionuclides moved through the pathway is evaluated using TDRW (Time Domain Random Walk) method for simulating the transport on the time-domain. Some retention mechanisms, such as radioactive decay, equilibrium sorption, and matrix diffusion, as well as the advection-dispersion were selected as the factors to influence on the elapsed time. From the simulation results, the effects of the sorption and matrix diffusion, determined by the properties of the radionuclides and underground media, on the transport of the radionuclides were analyzed and a decay chain of the radionuclides was also examined. The radionuclide ratio of the mass discharge into the surface environment to the mass released from the supposed repository did not exceed $10^{-3}$, and it decreased when the matrix diffusion were considered. The method used in this study could be used in preparing the data on radionuclide transport for a safety assessment of a geological disposal facility because the method could evaluate the travel time of the radionuclides considering the transport retention mechanism.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.63-71
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• 2011

A large volume of uranium electrokinetic leachate has been generated during the electrokinetic decontamination to remove uranium from contaminated soil. The treatment technology for the reuse of the uranium leachate was developed. The concentration of uranium in the generated uranium leachate was 180 ppm and concentrations of Mg(II), K(I), Fe(II), and Al(III) ions ranged from 20 ppm to 1,210 ppm. The treatment process for uranium leachate consisted mainly of mixing and cohesion, precipitation, concentration, and filtration. In order to obtain the pH=11 of a precipitate solution, the calcium hydroxide needs to be 3.0g/100ml and the sodium hydroxide needed to be 2.7g/100ml. The results of several precipitation experiments showed that a mixture of NaOH+0.2g alum+0.15g magnetite was an optimal precipitant for filtration. The average particle size of precipitate with NaOH+alum+0.15g magnetite was $600\;{\mu}m$. Because the total value of metal concentrations in supernatant at pH=9 was the smallest, sodium hydroxide should be added with 0.2g alum and 0.15g magnetite for pH=9 of leachate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.473-478
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• 2018

Radioactive aerosol generated in cutting and melting work during the NPP decommissioning process can cause internal exposure to body through workers' breath. Thus, it is necessary to assess worker internal exposure due to the radioactive aerosol during decommissioning. The actually measured value of the working environment is needed for accurate assessment of internal exposure, but if it is difficult to actually measure that value, the internal exposure dose can be estimated through recommended values such as the fraction of amount of intake and the size of particles suggested by the International Committee on Radiological Protection (ICRP). As for the selection of particle size, this study applied a value of $5{\mu}m$, which is the size of particles considering the worker recommended by the ICRP. As for the amount of generation, the amount of intake was estimated using data on the mass of aerosol generated in a melting facility at a site in Kozloduy, Bulgaria. In addition, using these data, this study calculated the level of radioactivity in the worker's body and stool and conducted an assessment of internal exposure using the BiDAS computer code. The internal exposure dose of Type M was 0.0341 mSv, that of Type S was 0.0909 mSv. The two types of absorption showed levels that were 0.17% and 0.45% of the domestic annual dose limit, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.255-263
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• 2006

Vol-oxidizer is a device to convert $UO_2$ pellets into $U_3O_8$ powder and to feed a homogeneous powder into a Metal Conversion Reactor in the ACP(Advanced Spent Fuel Conditioning Process). In this paper, we propose a design model of the vol-oxidizer, develop the new vol-oxidizer with a capacity of 20 kg HM/batch in $UO_2$ pellets, and conduct a verification for the device. Design considerations include the internal structure, the capacity, the heating position of the device, and the size. The dimensions of the new vol-oxidizer are decided by the design model. We determine a permeability test of the $U_3O_8$ measuring the temperature distribution, and the volume of $UO_2$ and $U_3O_8$. We manufactured the new vol-oxidizer for a 20 kg HM/batch in $UO_2$ pellets, and then analyzed the characteristics of the $U_3O_8$ powder for the verification. The experimental results show that the permeability of the $U_3O_8$ throughout mesh enhance more than old vol-oxidizer, the oxidation time takes only 8 hours when compared with the 13 hours of the old device, and the average distribution of particle size is $40{\mu}m$. The capacities of new vol-oxidizer for a 20 kg HM/batch in $UO_2$ pellets were agree well with the predictions of design model.