• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.35-43
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• 2015

A large amount of radioactively-contaminated gravel can be produced on the demolition/restoration of facilities related the back end of fuel cycle. However, because of the lacking in basic knowledge for decontamination of radioactive-contami-nated gravel, this study has performed the basic tests using for soil-washing. To find effective decontamination condition, several experiments were carried out for the selection of optimal decontamination agents. Washing by 0.1 M nitric acid was proved to be more effective than that by distilled water or surfactant for decontamination of uranium-contaminated gravel. In addition, crushing/grinding of uranium-contaminated gravel prior to washing was contributed to increase in of removal efficiency of uranium and reduction of decontamination time. The smaller the sizes of crushed gravel was, the more the removal efficiency increased. Also, small the sized particles improved chances for meeting the clearance requirement of the treated gravel.

• Journal of Korean Society of Environmental Engineers
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• v.35 no.12
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• pp.960-966
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• 2013

In this study, operation data from an actual plant in M city were analyzed to evaluate effects of water supply into sewage sludge on the co-incineration of municipal solid waste (MSW) and sewage sludge. Design capacity of the stoker incinerator is 50 m3/day. Maximum portion of sewage sludge in the total waste input was 20%. According to this research, moisture content increase up to 85% of sewage sludge could be possible by water supply on MSW input to the incinerator. Therefore, stable operation of incinerator could be achieved. As water was added into sewage sludge up to 85%, input quantity of sewage sludge to frequency (Hz) was similar to theoretical input. Also, it is concluded that stable management of incinerator without leachate combustion and damages of refractory bricks and castable could be achieved by the reduction of low heating value.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.1
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• pp.79-90
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• 2016

The second-stage near surface disposal facility for low and very low level radioactive waste's permanent disposal is to be built. During the institutional control period, the inadvertent intrusion of the general public is limited. But after the institutional control period, the access to the general public is not restricted. Therefore human who has purpose of residence and resource exploration can intrude the disposal facility. In this case, radioactive effects to the intruder should be limited within regulatory dose limits. This study conducted the safety assessment of human intrusion on the second-stage surface disposal facility through drilling and post drilling scenario. Results of drilling and post drilling scenario were satisfied with regulatory dose limits. The result showed that post-drilling scenario was more significant than drilling scenario. According to the human intrusion time and behavior after the closure of the facility, dominant radionuclide contributing to the intruder was different. Sensitivity analyses on the parameters about the human behavior were also satisfied with regulatory dose limits. Especially, manual redistribution factor was the most sensitive parameter on exposure dose. A loading plan of spent filter waste and dry active waste was more effective than a loading plan of spent filter waste and other wastes for the radiological point of view. These results can be expected to provide both robustness and defense in depth for the development of safety case further.

• Journal of the Korea Organic Resources Recycling Association
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• v.27 no.2
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• pp.5-12
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• 2019

The study is carried out to survey the proper management and to propose an eco - friendly separation system through efficient screening and resource recovery of excavated materials containing waste from various excavating fields such as reconstruction of landfill sites for reuse, reclamation of unsanitary landfill and residential land development of waste dumping sites. The current status and screening process and analytical characteristics of the excavated materials containing waste were reviewed. Through the analysis of the samples such as separated combustibles, recyclable soils and residues collected from the on-site visits we were able to understand the characteristics of separated materials and excavated materials containing waste such as calorific value, elementary composition, TOC, foreign material content and LOI. It has been found that elimination of the moisture of excavations, removal of attached soil from the surfaces of the excavated combustibles and the quantitative supply method of the input devices are the main operating factors as essential factors for the optimal separation of excavated materials containing waste. For efficient management and recycling of excavated materials containing, it is necessary to set criteria of ash content in separated combustibles and criteria organic matter content in separated soils.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.228-233
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• 2013

It has been continuously generated the requirement for the replacement of the main components such as a steam generator due to the deterioration of the nuclear power plant all around the world. Also, a large amount of radioactive metal was generated during the decommissioning in a short period. It is required to make an accurate measurement of the residual radioactivity for recycling the metal waste for releasing from regulatory control. In planning the measurement procedures, the influence of geometry, self-absorption, density and other relevant factors on the representativeness of the measurements should be considered for the decommissioning metal waste. In this study, the method for measurement procedures, the source term evaluation, the ways to secure representative samples, the measurement device for wide area and the self-absorption correction factors for different density were evaluated. The metal samples for measurement were prepared for securing the simple geometry and representative by melting process. The developed correction method for measuring the radioactivity a variety density of metal waste could improve the reliability of the evaluation results for clearance.

• 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.

• The Journal of Engineering Geology
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• v.21 no.4
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• pp.295-304
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• 2011

The first step in improving our understanding of uncertainties suclt as rock mass strength parameters and deformation modulus in rock masses around high-level radioactive waste disposal repositories, for improved safety, is to study the process of crack development in intact rock. Therefore, in this study, the fracture process and crack development were examined in samples of KURT granite taken from the KAERI Underground Research Tunnel (KURT), based on acoustic emission (AE) and moment tensor analysis. The results show that crack initiation, coalescence, and unstable crack occurred at rock uniaxial compressive strengths of 0.45, 0.73, and 0.84, respectively. In addition, moment tensor analysis indicated that during the early stage of loading, tensile cracks were predominant. With increasing applied stress, the number of shear cracks gradually increased. When the applied stress exceeded the stress level required for crack damage, unstable shear cracks which directly result in failure of the rock were generated along the failure plane.

• Tunnel and Underground Space
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• v.34 no.4
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• pp.312-329
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• 2024

In various underground research projects such as energy storage and development and radioactive waste disposal targeting deep underground, the characteristics of permeable rock fractures that serve as major pathway of groundwater flow in deep rock aquifer are considered as an important evaluation factor in the design, construction, and operation of research facilities. In Korea, there is little research and database on the location and hydraulic characteristics of permeable rock fractures and the pattern of groundwater flow patterns that may occur between fractures in deep rock boreholes. In this paper, the hydraulic characteristics of permeable rock fractures in deep rock aquifer were evaluated through the analysis of geothermal gradient and pumping test data. First, the deep geothermal distribution was identified through temperature logging, and the geothermal gradient was obtained through linear regression analysis using temperature data by depth. In addition, the hydraulic characteristics of the fractured rock were analyzed using outflow temperature obtained from pumping tests. Ultimately, the potential location and hydraulic characteristics of permeable rock fractures, as well as groundwater flow within the boreholes, were evaluated by integrating and analyzing the geophysical logging and hydraulic testing data. The process and results of the evaluation of deep permeable rock fractures proposed in this study are expected to serve as foundational data for the successful implementation of underground research projects targeting deep rock aquifers.

• Journal of Radiation Industry
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• v.17 no.4
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• pp.359-367
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• 2023

In this study, the function and purpose of the disposal cover, which is an engineering barrier installed to isolate the disposal vault of the near-surface disposal facility for radioactive waste from natural/man-made intrusion, and the design details of the demonstration facility for performance verification were described. The Demonstration facility was designed in a partially divided form to secure the efficiency of measurement while being the same as the actual size of the surface disposal facility to be built in the Intermediate & low-level radioactive waste disposal site of the Korea Radioactive Waste Agency (KORAD). The instruments used for measurement consist of a multi-point thermometer, FDR (Frequency Domain Reflectometry) sensor, inclinometer, acoustic sensor, flow meter, and meteorological observer. It is used as input data for the monitoring system. The 3D monitoring system was composed of 5 layers using the e-government standard framework, and was developed based on 4 components: screen, control module, service module, and DBIO(DataBase Input Output) module, and connected them to system operation. The monitoring system can provide real-time information on physical changes in the demonstration facility through the collection, analysis, storage, and visualization processes.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.183-193
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• 2018

Currently, the Korea Atomic Energy Research Institute (KAERI) is planning to build the Ki-Jang Research Reactor (KJRR) in Ki-Jang, Busan. It is important to safely dispose of low-level radioactive waste from the operation of the reactor. The most efficient way to treat radioactive waste is cement solidification. For a radioactive waste disposal facility, cement solidification is performed based on specific waste acceptance criteria such as compressive strength, free-standing water, immersion and leaching tests. Above all, the leaching test is important to final disposal. The leakage of radioactive waste such as $^{137}Cs$ causes not only regional problems but also serious global ones. The cement solidification method is simple, and cheaper than other solidification methods, but has a lower leaching resistance. Thus, this study was focused on the development of cement solidification for an enhancement of cesium leaching resistance. We used Zeolite and Loess to improve the cesium leaching resistance of KJRR cement solidification containing simulated KJRR liquid waste. Based on an SEM-EDS spectrum analysis, we confirmed that Zeolite and Loess successfully isolated KJRR cement solidification. A leaching test was carried out according to the ANS 16.1 test method. The ANS 16.1 test is performed to analyze cesium ion concentration in leachate of KJRR cement for 90 days. Thus, a leaching test was carried out using simulated KJRR liquid waste containing $3000mg{\cdot}L^{-1}$ of cesium for 90 days. KJRR cement solidification with Zeolite and Loess led to cesium leaching resistance values that were 27.90% and 21.08% higher than the control values. In addition, in several tests such as free-standing water, compressive strength, immersion, and leaching tests, all KJRR cement solidification met the waste acceptance or satisfied the waste acceptance criteria for final disposal.