• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.257-265
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• 2007

Spent filter wastes of about 1,000 units (200 L) have been stored in the waste storage facility of the Korea Atomic Energy Research Institute since its operation. At the moment, to secure space in a waste storage facility as well as to efficiently manage spent filter wastes, it is necessary to conduct a compaction treatment of these spent filters, and finally, to repack the compacted spent filters into a 200 liter drum. To do that, the spent filter wastes were first classified according to their generation facilities, their generation date and their surface dose rate by investigating the inventory of the spent filters. In order to repack a compacted spent filter in a 200 liter drum, it is first necessary to conduct a radionuclide assessment of a spent filter before compacting it. Therefore, after taking a representative sample from a spent filter without a dismantlement, the nuclide analysis for it will be conducted. And then, after putting a spent filter into a regular drum by conducting the columnar shaping of the hexahedral form of a spent filter, the compaction treatment of the shaped spent filter will be conducted by vertically compacting it.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.23-29
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• 2013

LCC (Liquid cadmium cathode) is used for electrowinning in pyroprocessing to recover uranium and transuranic elements simultaneously. It is one of the core technologies in pyroprocessing with higher proliferation resistance than a wet reprocessing because LCC-cell does not separate TRU from uranium. The crucible which holds the LCC is technically important because it should be nonconducting material to prevent deposition of metallic elements on the crucible outer surface. The chemical stability is also crucial factor to choose crucible material due to the strong reactivities of TRU and possible incorporation of Li metal during the operation. In this study, the chemical stabilities of four kinds of representative ceramic materials such as $Al_2O_3$, MgO, $Yl_2O_3$ and BeO were thermodynamically and experimentally evaluated at $500^{\circ}C$ with simulated LCC. The contact angle of LCC on ceramic materials was measured as function of time to predict chemical reactivity. $All_2O_3$ showed poorest chemical stability and the pores in BeO contributed to a decreases in contact angle. MgO and $Y_2O_3$ have superior chemical stability among the materials.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.1-9
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• 2013

The rare earth oxide wastes consisting of major 8 nuclides Y, La, Ce, Pr, Nd, Sm, Eu and Gd, are generated during the salt waste treatment of PyroGreen process. The final form of the rare earth is generated as the oxide state. In this study, six candidate glasses were developed to evaluate the feasibility for vitrifying the rare earth oxide wastes within the borosilicate glass system. The solubilities of the mixture of the rare earth oxide waste showed less than 25wt% at $1,200^{\circ}C$, less than 30wt% at $1,300^{\circ}C$, respectively. It means that solubility is increased with the temperature increment. The liquidus temperature of the homogeneous glass with 20wt% waste loading was determined as less than $950^{\circ}C$. In more than solubility of rare earth oxides glass, formation of rare earth-oxide-silicate crystal in glass-ceramic occurred as the secondary phase. As their viscosity at melting temperature $1,200{\sim}1,300^{\circ}C$ was less than 100 poise, electrical conductivity was higher than 1 S/cm, 20~25wt% waste loading glasses with good surface homogeneity are judged to have good operability in cold crucible induction melter. Other physicochemical properties of the developed glasses are going to be experimented in the future.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.161-172
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• 2017

To evaluate radiological impact from the operation of a low- and intermediate-level radioactive waste disposal facility, a logical presentation and explanation of expected accidental scenarios is essential to the stakeholders of the disposal facility. The logical assessment platform and procedure, including analysis of the safety function of disposal components, operational hazard analysis, operational risk analysis, and preparedness of remedial measures for operational safety, are improved in this study. In the operational risk analysis, both design measures and management measures are suggested to make it possible to connect among design, operation, and safety assessment within the same assessment platform. For the preparedness of logical assessment procedure, classification logic of an operational accident is suggested based on the probability of occurrence and consequences of assessment results. The improved assessment platform and procedure are applied to an operational accident analysis of the Korean low- and intermediate-level radioactive waste disposal facility and partly presented in this paper.

• Journal of Energy Engineering
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• v.20 no.3
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• pp.191-199
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• 2011

The char burn-out and NOx emissions from sub-bituminous coal were investigated in drop tube furnace under $O_2/N_2$ and $O_2/CO_2$ environments with different $O_2$ concentrations of 12, 21 and 31%. Results show that the char burn-out rate is faster as $O_2$ concentration increases higher and char burn-out rate under $O_2/CO_2$ decreases due to the lower oxygen diffusion into coal surface through the $CO_2$ rich boundary layer. NO concentration increases with increasing $O_2$ concentration, but declines at $O_2$ concentration of 31%. Meanwhile, NO emission indexes decreases monotonically with increasing $O_2$ concentration, which indicates that more NO reduction occurs with higher $O_2$ concentration probably due to greater HCN formation. For all conditions of $O_2$ concentration, the NO concentration under $O_2/N_2$ maintains higher than those of $O_2/CO_2$ due to presence of thermal NO.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.235-243
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• 2006

In this study, an experimental study on the sorption properties of uranium(VI) onto a bentonite colloid generated from Gyeongju bentonite which is a potential buffer material in a high-level radioactive waste repository was performed as a function of the pH and the ionic strength. The bentonite colloid prepared by separating a colloidal fraction was mainly composed of montmorillonite. The concentration and the size fraction of the prepared bentonite colloid measured using a gravitational filtration method was about 5100 ppm and 200-450 nm in diameter, respectively. The amount of uranium removed by the sorption reaction bottle walls, by precipitation, and by ultrafiltration was analyzed by carrying out some blank tests. The removed amount of uranium was found not to be significant except the case of ultrafiltration at 0.001 M $NaClO_4$. The ultrafiltration was significant in the lower ionic strength of 0.001 M $NaClO_4$ due to the cationic sorption onto the ultrafilter by a surface charge reversion. The distribution coefficient $K_d$ (or pseudo-colloid formation constant) of uranium(VI) for the bentonite colloid was about $10^4{\sim}10^7mL/g$ depending upon pH and ionic strength of $NaClO_4$ and the $K_d$ was highest in the neutral pH around 6.5. It is noted that the sorption of uranium(VI) onto the bentonite colloid is closely related with aqueous species of uranium depending upon geochemical parameters such as pH, ionic strength, and carbonate concentration. As a consequence, the bentonite colloids generated from a bentonite buffer can mobilize the uranium(VI) as a colloidal form through geological media due to their high sorption capacity.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.217-226
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• 2006

The structural and contamination characteristics of hot cells at KAERI were investigated. The SEM results showed that the size of the hot particulate on the inner surface of the hot cell ranged from 0.2 to $10{\mu}m$. It was found that an inlet flow rate of 15 m/sec was suitable for this developed cyclone with a 49 mm optimum vortex finder length. The results showed that the collection efficiency was about 85% for $3{\mu}m$ particles. The collection efficiency didn't show a sharp increase when the inlet flow rate was faster than 15m/sec. When the temperature of the inlet flow gas was increased, the collection efficiency of the cyclone was slightly decreased. The larger the vortex finder length was, the higher the pressure drop in the cyclone was. The cut size diameter decreased with an increment of the Reynolds number. It was established that the flow in the cyclone was a turbulent flow on the basis of the Reynolds number and this turbulent flow caused a pressure drop in the cyclone. $Stk^{1/2}_{50}$ decreased with increasing values of the Reynolds number and it gradually approached a constant value at a higher value of the Reynolds number Namely, $Stk^{1/2}_{50}$ approached approximately 0.045 between 6000 and 8000 of the Reynolds number.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.73-86
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• 2015

The use of complimentary indicators, other than radiation dose and risk, to assess the safety of radioactive waste disposal has been discussed in a number of publications for providing the reasonable assurance of disposal safety and convincing the public audience. In this study, the radionuclide flux was selected as performance indicator to appraise the performance of engineered barriers and natural barrier in the Wolsong low- and intermediate-level waste disposal facility. Radionuclide flux showing the retention capability by each compartment of the disposal system is independent of assumptions in biosphere model and exposure pathways. The scenario considered as the normal scenario of disposal facility has been divided into intact or degraded silo concrete conditions. In the intact silo concrete, the radionuclide flux has been assessed with respect to the radionuclide retardation performance of each engineered barrier. In the degraded silo concrete, the radionuclide flux has been explored based on the performance degradation of engineered barriers and the relative significance of natural barrier quantitatively. The results can be used to optimally design the near-surface disposal facility being planned as the second project phase. In the future, additional complimentary indicators will be employed for strengthening the safety case for improving the public acceptance of low- and intermediate-level waste disposal facility.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.37-43
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• 2012

Radioactive waste transport package was developed to transport eight drums of low and intermediate level waste(LILW) in accordance with the IAEA and domestic related regulations. The package is classified with industrial package IP-2. IP-2 package is required to undergo a free drop test and a stacking test. After free drop and stacking tests, it should prevent the loss or dispersal of radioactive contents, and loss of shielding integrity which would result in more than 20 % increase in the radiation level at any external surface of the package. The objective of this study is to establish the safety test method and procedure for stacking test and to prove the structural integrities of the IP-2 package. Stacking test and analysis were performed with a compressive load equal to five times the weight of the package for a period of 24 hours using a full scale model. Strains and displacements were measured at the corner fitting of the package during the stacking test. The measured strains and displacements were compared with the analysis results, and there were good agreements. It is very difficult to measure the deflection at the container base, so the maximum deflection of the container base was calculated by the analysis method. The maximum displacement at the corner fitting and deflection at the container base were less than their allowable values. Dimensions of the test model, thickness of shielding material and bolt torque were measured before and after the stacking test. Throughout the stacking test, it was found that there were no loss or dispersal of radioactive contents and no loss of shielding integrity. Thus, the package was shown to comply with the requirements to maintain structural integrity under the stacking condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.143-149
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• 2012

Groundwater flow simulations were performed to obtain data of groundwater flow used in a safety assessment for a hypothetical geological disposal facility assumed to be located in the KURT (KAERI Underground Research Tunnel) site. A regional scale modeling of the groundwater flow system was carried out to make boundary conditions for a local scale modeling. And, fracture zones identified at the study site were involved in the local scale groundwater flow model. From the results of the local scale modeling, a hydraulic head distribution was indicated and it was used in a particle tracking simulation for searching pathway of groundwater from the location of the hypothetical disposal facility to the surface where the groundwater reached. The flow distance and discharge rate of the groundwater in the KURT site were calculated. It was thought that the modeling methods used in this study was available to prepare the data of groundwater flow in a safety assessment for a geological disposal facility of radioactive wastes.