• Journal of Korean Society of Water and Wastewater
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• v.28 no.1
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• pp.83-89
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• 2014

The aim of this study is to investigate the sorption/ion exchange of radioactive nuclides such as $Cs^+$ and $Sr^{2+}$ by synthetic Na-micas. In order to prepare Na-micas, two natural micas (phlogopite and biotite) were used as precursor materials. XRD, SEM, and EDS analyses were used to examine material characterization of synthetic Na-micas. Analyses of materials revealed that Na-micas were successfully obtained from natrual micas by K removal treatment. On the other hand, single solute (Cs or Sr) and bi-solute (Cs/Sr) sorption experiments were carried out to determine sorption capacity of Na-micas for Cs and Sr under different pH and ionic strength conditions. Uptake of Cs and Sr by micas in bi-solute system was lower than in single-solute system. Additionally, Langmuir and Langmuir competitive models were applied to describe sorption isotherm of Na-micas. bi-solute system was well described by Langmuir competitive models. For the results obtained in this study, Na-micas could be promising sorbents to treat multi-radioactive species from water and groundwater.

• Journal of Radiation Protection and Research
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• v.13 no.2
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• pp.52-56
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• 1988

An ion-exchange method for the detection of radioactivity in water using ion-exchange resin in concentrating radioactive nuclides was compared with an evaporation method. The loss of the radioactive materials in the sample treated by the ion-exchange method was less by about 20% than that by the evaporation method. In addition, the evaporation method needed about 20 hours for evaporating one liter of the sample at $70^{\circ}C$, while the ion-exchange method spent 6 hours to adsorb and adsorb the same amount of the sample on the resin. Consequently, the ion-exchange method is more effective than the evaporation method for the treatment of the radioactively contaminated water and is especially suitable for detecting the low-level radioactivity in water.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.4
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• pp.479-489
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• 2021

The Republic of Korea is expected to participate in the denuclearization verification activities by the International Atomic Energy Agency (IAEA) in case any neighboring countries declared denuclearization. In this study, samples for the verification of nuclear activities in undeclared areas were selected for the denuclearization of neighboring countries, and the appropriateness of the procedures was considered. If a country with nuclear weapons declares denuclearization, it must be accompanied by the IAEA's verification regarding nuclear materials and weapons in the declared and undeclared areas. The analysis of the process samples or on-site environmental samples and the verification of undeclared nuclear facilities and materials aid in uncovering any evidence of concealment of nuclear activity in undeclared areas. Therefore, a methodology was established for effective sampling and analysis in accordance with proper procedures. Preparations for sampling in undeclared areas were undertaken for various potential scenarios, such as, the establishment of zones according to radiation dose, methods of supplying electricity, wireless communication networks, targets of sampling according to characteristics of nuclides, manned sampling method, and unmanned sampling method. Through this, procedures were established for pre- and post-site settings in preparation for hazards and limiting factors at nuclear inspection sites.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.349-356
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• 2022

Nuclear power plant decommissioning generates significant concrete waste, which is slightly contaminated, and expected to be classified as clearance concrete waste. Clearance concrete waste is generally crushed into rubble at the site or a satellite treatment facility for practical disposal purposes. During the process, workers are exposed to radiation from the nuclides in concrete waste. The treatment processes consist of concrete cutting/crushing, transportation, and loading/unloading. Workers' radiation exposure during the process was systematically studied. A shielding package comprising a cylindrical and hexahedron structure was considered to reduce workers' radiation exposure, and improved the treatment process's efficiency. The shielding package's effect on workers' radiation exposure during the cutting and crushing process was also studied. The calculated annual radiation exposure of concrete treatment workers was below 1 mSv, which is the annual radiation exposure limit for members of the public. It was also found that workers involved in cutting and crushing were exposed the most.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.33-40
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• 2006

The characteristics of the aluminum waste melting and the distribution of the radioactive nuclides have been investigated for the estimation on the volume reduction and the decontamination of the aluminum wastes from the decommissioning of the TRIGA MARK it and III research reactors at the Korea Atomic Energy Research Institute(KAERI). The aluminum wastes were melted with the use of the fluxes such as flux $A:NaCl-KCl-Na_3AlF_6$, flux B:NaCl-NaF-KF, flux $C:CaF_2$, and flux $D:LiF-KCl-BaCl_2$ in the DC graphite arc furnace. For the assessment of the distribution of the radioactive nuclides during the melting of the aluminum, the aluminum materials were contaminated by the surrogate nuclides such as cobalt(Co), cesium(Cs) and strontium(Sr). The fluidity of aluminum melt was increased with the addition of the fluxes, which has slight difference according to the type of fluxes. The formation of the slag during the aluminum melting added the flux type C and D was larger than that with the flux A and B. The rate of the slag formation linearly increased with increasing the flux concentration. The results of the XRD analysis showed that the surrogate nuclide was transferred to the slag, which can be easily separated from the melt and then they combined with aluminum oxide to form a more stable compound. The distribution ratio of cobalt in ingot to that in slag was more than 40% at all types of fluxes. Since vapor pressures of cesium and strontium were higher than those that of the host metals at the melting temperature, their removal efficiency from the ingot phase to the slag and the dust phase was by up to 98%.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.971-976
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• 2012

Silica-coated Au with Ag, Co, Cu, and Ir bimetallic radioisotope nanoparticles were synthesized by neutron irradiation, after coating $SiO_2$ onto the bimetallic particles by the sol-gel St$\ddot{o}$ber process. Bimetallic nanoparticles were synthesized by irradiating aqueous bimetallic ions at room temperature. Their shell and core diameters were recorded by TEM to be 100 - 112 nm and 20 - 50 nm, respectively. The bimetallic radioisotope nanoparticles' gamma spectra showed that they each contained two gamma-emitting nuclides. The nanoparticles could be used as radiotracers in petrochemical and refinery processes that involve temperatures that would decompose conventional organic radioactive labels.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.879-884
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• 1995

The results of performance assessment analyses have shown that the long-lived radionuclides such as I-129 control the potential individual dose impact to the public. I-129 is difficult-to-measure(DTM) in low-level waste because it is non-gamma emitting radionuclides and exists at extremely low concentrations in radioactive waste generated by nuclear reactors. In this study, computer modeling technique to predict release rate of I-129 is developed to provide another tools far performance assessment of land disposal facilities and characteristics of radwaste. Model suggested in this study will give conservative values of I-129 release rate far determination of radwaste characteristics. More detailed approach is implemented to account for release conditions of fuel source-nuclides. 1-131 concentration measured from reactor coolant and released fraction from tramp fuel have dominant roles in calculating release rate of I-129 with fuel defect conditions.

• Journal of the Korean Society of Radiology
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• v.12 no.5
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• pp.565-573
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• 2018

Cyclotron for medical purposes generates nuclear reaction by accelerating protons in high speed, in order to produce radiopharmaceuticals, and unnecessary neutrons are generated through such nuclear reaction. Neutrons cause activation in the parts of cyclotron which then cause exposure to radiation for people working in the field. This study, in that regard, aims to analyze exposure level by finding out the degree of activation of aluminum body, silver body, and havar foil which are the parts of Targetry where the nuclear reaction takes place. The results of the experiment showed that aluminum body and silver body had no problems re-using them as the energy and half-life of activated nuclides were small and short, making the affect on the people working in the field extremely low. However for havar foil, its activated nuclides had a high level of energy which resulted in high level of affect to the people working in the field. The activation factors of the cyclotron were analyzed, and the results showed that the Havar foil was activated the most among the targetry parts, and greatly exposed workers due to regular replacement, and needed special management as radioactive waste.

• Tunnel and Underground Space
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• v.11 no.2
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• pp.132-145
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• 2001

The purpose and need of the study is to quantify the advantage or disadvantage of the environmental friendliness of the partitioning of nuclear fuel cycle. To this end, a preliminary study on the quantitative effect of the partition on the permanent disposal of spent PWR and CANDU fuel (HLW) was carried out. Before any analysis, the so-called reference radionuclide release scenario from a potential repository embedded into a crystalline rock was developed. Firstly, the feature, event and processes (FEPs) which lead to the release of nuclides from waste disposed of in a repository and the transport to and through the biosphere were identified. Based on the selected FEPs, the ‘Well Scenario’which might be the worst case scenario was set up. For the given scenario, annual individual doses to a local resident exposed to radioactive hazard were estimated and compared to that from direct disposal. Even though partitioning and transmutation could be an ideal solution to reduce the inventory which eventually decreases the release time as well as the peaks in the annual dose and also minimize the repository area through the proper handling of nuclides, it should overcome major disadvantages such as echnical issues on the partitioning and transmutation system, cost, and public acceptance, and environment friendly issues. In this regard, some relevant issues are also discussed to show the direction for further studies.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.73-78
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• 2009

The decay heat of Cs and Sr contained in a LiCl waste salt, generated from an electrolytic reduction process in pyroprocessing of spent nuclear fuel, has been calculated. The calculation has been carried out under some assumptions that most of the LiCl waste is purified and recycled to main process, and the residual is fabricated to make a waste form. As a result, the decay heat from daughter nuclides such as Ba and Y seems to be maximum 4.6 times higher than that from their parent nuclides such as Cs and Sr. The thermal release from Cs and Sr in the LiCl waste is the maximum around the first one month, so an cooling system operation for some time at the beginning would be suggested to control a rapid increase in the temperature of the LiCl waste salt.