• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.303-313
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• 2023

Spent fuels (SFs) are stored in a storage pool after discharge from nuclear power plants. They can be transferred to for the further processes such as dry storage sites, processing plants, or disposal sites. One of important measures of SF is the burnup. Since the radioactivity of SF is strongly dependent on its burnup, the burnup of SF should be well estimated for the safe management, storage, and final disposal. Published papers about the methodology for the burnup estimation from the known activities of important radioactive sources are somewhat rare. In this study, we analyzed the dependency of the burnup on the important radiation source activities using ORIGEN-ARP, and suggested simple correlations that relate the burnup and the important source activities directly. A burnup estimation equation is suggested for PWR fuels relating burnup with total neutron source intensity (TNSI), initial enrichment, and cooling time. And three burnup estimation equations for major gamma sources, ¹³⁷Cs, ¹³⁴Cs, and ¹⁵⁴Eu are also suggested.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.281-299
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• 2017

The inventory management of radionuclides is essential for the safe management of disposal facilities. In this study, we compared the activity of dry active waste predicted using the generated waste data and that measured for the accepted waste in the disposal facility. For very low level waste, the measured activity was higher than the predicted activity for $^{137}CS$, $^{90}Sr$, $^{99}Tc$ and $^{129}I$. In low level waste, the predicted activity was higher than the measured activity for all radionuclides. We also evaluated the variation in the predicted quantity and total activity of each level of dry active waste through a sensitivity analysis on a scaling factor. This result will contribute to the construction of a Safety Case and safe operation of disposal facilities.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.115-122
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• 2010

In this study, the removal of Cs and Tc from a simulated fission products (FP) solution which were co-dissolved with U during the oxidative-dissolution of spent fuel in a mixed carbonate solution of $(Na_2CO_3-NaHCO_3)-H_2O_2$ was investigated by using a selective precipitation method. As Cs and Tc might cause an unstable behavior due to the high decay heat emission of Cs as well as the fast migration of Tc when disposed of underground, it is one of the important issues to removal them in views of the increase of disposal safety. The precipitation of Cs and Re (as a surrogate for Tc) was examined by introducing sodium tetraphenylborate (NaTPB) and tetraphenylphosponium chloride (TPPCl), respectively. Precipitation of Cs by NaTPB and that of Re by TPPCl were completed within 5 minutes. Their precipitation rates were not influenced so much by the temperature and stirring speed even if they were increased by up to $50^{\circ}C$ and 1,000 rpm. However, the pH of the solution was found to have a great influence on the precipitation with NaTPB and TPPCl. Since Mo tends to co-precipitate with Re at a lower pH, especially, it was effective that a selective precipitation of Re by TPPCl was carried out at pH of above 9 without co-precipitation of Mo and Re. Over 99% of Cs was precipitated when the ratio of [NaTPB]/[Cs]>1 and more than 99% of Re, likewise, was precipitated when the ratio of [TPPCl]/[Re]>1.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.231-241
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• 2022

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as ¹²⁹I and ¹³⁵Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM^-1.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2022.11a
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• pp.167-168
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• 2022

Recently, starting with the permanent suspension of Gori 1 in Korea, the importance of the disposal of concrete structures in nuclear power plants has emerged, and environmental and safety are required to be proved accordingly. Safe radioactive waste disposal technology that immobilizes harmful radioactive elements, which are by-products of nuclear power, inside a solid matrix and recycling measures are needed to secure an efficient waste disposal space. This study was conducted to confirm whether recycled cement generated in the process of radioactive concrete treatment can be used as a solidifying material for radioactive waste treatment. In order to simulate the concrete exposed to radiation, aqueous solutions of Di-water, CsCl 1M, and CoCl₂ 1M were used as blending water at W/B 0.5. Tricalcium phosphate and Prussian blue were substituted with 5 wt.% based on the weight of recycled cement as a binder to improve solidification performance, and their hydration characteristic was analyzed.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2023.05a
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• pp.305-306
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• 2023

Recently, as the radioactive waste disposal facility becomes scarce, the importance of efficient disposal of waste from nuclear power plants is increasing. This study was conducted to utilize radioactive waste concrete powder as solidifying agent for radioactive waste treatment. Paste with an age of more than one year was used with a disk mill to have a particle size of 150㎛ or less, and treated at temperatures of 500℃, 600℃ and 700℃ for 2 hours. In order to simulate the radioactive cement powder, aqueous solutions of Di-water, CsCl 1M, SrCl₂ 1M and CoCl₂ 1M were used as blending water at W/C 0.7 and to improve fluidity, polycarboxylate type superplasticizer was used at 0.4 wt.% based on the weight of recycled cement paste powder. Characterisation was carried out using vicat method, strength and density.

• Nuclear Engineering and Technology
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• v.16 no.2
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• pp.64-69
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• 1984

With regard to the radioactive waste disposal, adsorption properties and migration rates have been evaluated for Cs-137 and Sr-90 with the domestic clay sampled from Cnyang, Sanchong and Mooan. Sorption coefficients (Ksorp) were determined by batch experiments. The measured values of Ksorp were ranged from 8000 to 17,000 ml/gr for Cs-137 of 0.1$\mu$Ci/ml, and from 10,000 to 15,000m1/gr for Sr-90 of 0.l$\mu$Ci/ml. Remarkably, Mooan clay showed lower values of Ksorp than those of the others. This could be explained by the poor soprtion capacity of the quartz found only in the Mooan clay. For the quantitative analysis, sorption isotherm equations of Freundlich type were made with the obtained values of Ksorp. $C_{R}$=18.0 $C_{A}$$^{0.74}$ : Cs-137, $C_{R}$=0.84 $C_{A}$$^{0.45}$ : Sr-90. By introducing the BOX model combined with the above relationships, simulation of underground nuclide movement was carried out. The results showed that the domestic clays could be the effective backfill material for repositories.itories.ies.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.113-121
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• 1985

For the safety assessment of radwaste treatment and disposal, the leaching characteristics of $Cs^+\;and\;Sr^{++}$ from the cement-solidified radwastes was investigated by means of the survey and analysis of the activity leach rate and cumulative leach fraction of $^{137}Cs\;and\;^{90}Sr$ leaching from solidified radwastes. The cylindrical cement-solidified samples were made by the sodium silicate method and neutralization method changing the mole density of simulated liquid radwaste, type of cement, adding material and concentration of boric acid. The leaching study was carried out by the method using simulated PWR concentrated liquid radwaste solidified in cement that had been processed following the recommendations of IAEA. All the experimental results are in well accordance with the research data reported previously, but the watertight cement-solidified radwastes show rather high leach rate of radioactivity. It is learned, therefore, that the watertight cement is not adequate as the material for the solidified radwastes.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.233-238
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• 2003

Several types of gamma-ray sensors were constructed by packing different numbers of fibers into aluminum tubes, and tested to detect the $^137Cs$ gamma ray. It has been found from this investigation that tapered fibers can be more efficient to collect the lights produced inside the sensor and transfer them into the transmitting fiber in order to predict the functioning of the tapered fiber, tapered plastic scintillators, composed of polystyrene with minute amount of dPOPOP and PPO or dPBD, were fabricated and tested for the detection of gamma rays from 1.0 1.5 3.0 5.0 ${\mu}Ci$ $^137Cs$ sources, and the pulse hight spectrum and the relationship between the radioactivity and the total counts are analyzed. It has been found from this study that the tapered scintillating optical fiber, if manufactured, can be practically applied to the development of gamma-ray sensors which can be deployed In ${\mu}Ci$-level radiation fields

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.271-281
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• 2023

A study was conducted on the vitrification of the rare earth oxide waste generated from the PyroGreen process. The target rare earth waste consisted of eight elements: Nd, Ce, La, Pr, Sm, Y, Gd, and Eu. The waste loading of the rare earth waste in the developed borosilicate glass system was 20wt%. The fabricated glass, processed at 1,200℃, exhibited uniform and homogeneous surface without any crystallization and precipitation. The viscosity and electrical conductivity of the melted glass at 1,200℃ were 7.2 poise and 1.1 S·cm^-1, respectively, that were suitable for the operation of the vitrification facility. The calculated leaching index of Cs, Co, and Sr were 10.4, 10.6, and 9.8, respectively. The evaluated Product Consistency Test (PCT) normalized release of the glass indicated that the glass satisfied the requirements for the disposal acceptance criteria. Furthermore, the pristine, 90 days water immersed, 30 thermal cycled, and 10 MGy gamma ray irradiated glasses exhibited good compressive strength. The results indicated that the fabricated glass containing rare earth waste from the PyroGreen process was acceptable for the disposal in the repository, in terms of chemical durability and mechanical strength.