• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.563-575
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• 2020

The purpose of this study is to evaluate the activation characteristics that occur in a linear accelerator for container security inspection. In the computer simulation design, first, the targets consisted of a tungsten (Z=74) single material target and a tungsten (Z=74) and copper (Z=29) composite target. Second, the fan beam collimator was composed of a single material of lead (Z=82) and a composite material of tungsten (Z-74) and lead (Z=82) depending on the material. Final, the concrete in the room where the linear accelerator was located contained magnetite type and impurities. In the research method, first, the optical neutron flux was calculated using the MCNP6 code as a F4 Tally for the linear accelerator and structure. Second, the photoneutron flux calculated from the MCNP6 code was applied to FISPACT-II to evaluate the activation product. Final, the decommissioning evaluation was conducted through the specific activity of the activation product. As a result, first, it was the most common in photoneutron targets, followed by a collimator and a concrete 10 cm deep. Second, activation products were produced as by-products of W-181 in tungsten targets and collimator, and Co-60, Ni-63, Cs-134, Eu-152, Eu-154 nuclides in impurity-containing concrete. Final, it was found that the tungsten target satisfies the permissible concentration for self-disposal after 90 days upon decommissioning. These results could be confirmed that the photoneutron yield and degree of activation at 9 MeV energy were insignificant. However, it is thought that W-181 generated from the tungsten target and collimator of the linear accelerator may affect the exposure when disassembled for repair. Therefore, this study presents basic data on the management of activated parts of a linear accelerator for container security inspection. In addition, When decommissioning the linear accelerator for container security inspection, it is expected that it can be used to prove the standard that permissible concentration of self-disposal.

• Korean Journal of Mineralogy and Petrology
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• v.35 no.1
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• pp.25-39
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• 2022

This study was conducted to evaluate the manufacturing process of non-sintered cement for the safe containment of radioactive waste using low level or ultra-low level radioactive waste soil generated from nuclear-decommissioning facilities, clay minerals, and blast furnace slag (BFS) as an industrial by-product recycling and to characterize the products using mineralogical and morphological analyses. A stepwise approach was used: (1) measuring properties of source materials (reactants), such as waste soil, clay minerals, and BFS, (2) manufacturing the non-sintered cement for the containment of radioactive waste using source materials and deducing the optimal mixing ratio of solidifying and adjusting agents, and (3) conducting mineralogical and morphological analyses of products from the hydration reactions of manufactured non-sintered cement solidifier (NSCS) containing waste concrete generated from nuclear-decommissioning facilities. The analytical results of NSCS using waste soil and clay minerals confirmed none of the hydration products, but calcium silicate (CSH) and ettringite were examined as hydration products in the case of using BFS. The compressive strength of NSCS manufactured with the optimum mixing ratio and using waste soil and clay minerals was 3 MPa after the 28-day curing period, and it was not satisfied with the acceptance criteria (3.44 MPa) for being brought in disposal sites. However, the compressive strength of NSCS using BFS was estimated to be satisfied with the acceptance criteria, despite manufacturing conditions, and it was maximized to 27 MPa at the optimal mixing ratio. The results indicate that the most relevant NSCS for the safe containment of radioactive waste can be manufactured using BFS as solidifying agent and using waste soil and clay minerals as adsorbents for radioactive nuclides.

• Journal of the Korea institute for structural maintenance and inspection
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• v.17 no.3
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• pp.10-19
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• 2013

Recently, researches related to precast modular construction have been actively conducted for nuclear power plant, LNG gas tank, and small-medium PCCV as well as bridges and buildings. In this study, the precast panel cast with steel mesh reinforced mortar (SRM) which is similar reinforced ferrocement was developed for efficient precast construction, construction time reduction, and easy transportation. Mortar mixture with high strength and flowability was obtained from various case studies using silica fume and GGBS. Also, $1,200{\times}600{\times}150mm$ SRM and reinforced concrete (RC) panels were manufactured with reinforcing ratio of 2% and 4%. To verify structural performance of the SRM specimen, the basic material tests, free shrinkage test, and 3-point flexural test with a line loading were carried out. From the test results, it was determined that SRM specimens showed outstanding flexural capacity and ductility. However, the 4% reinforced SRM specimens must consider shear reinforcing to be used as a precast modular member.

• Korean Journal of Construction Engineering and Management
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• v.15 no.5
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• pp.138-146
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• 2014

Modular construction is one of the arising issues in the Korean construction industry due to shorter construction period and higher productivity on site. It is also expected that modular construction market would expand rapidly because of increasing demand of small houses by nuclear families and eco-friendly construction. However modular construction cases have been limited in Korea because of little demand in the market and the construction prices are even higher that the regular reinforced concrete methods. Even worse, because there is no standard bill of services exists in the industry, it may lead to the risk of construction expenses. So, the purpose of this study is focused on the development of bill of services which reflects characteristics and process of modular construction. In order to do this, the concepts and process of modular construction, WBS and CBS are analyzed. Analyses of the case studies and experts interview were executed so that proper principles for organizing bill of service of modular construction is established. Finally according to the 3 steps of modular construction, factory manufacturing, on-site modular construction and on-site general construction, the standard bill of services were suggested with the related coding system.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.28 no.6
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• pp.715-722
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• 2015

The prestressing force has not been managed after construction nevertheless it is one of the importrant factors that maintain the structural safety of PSC girder bridges. The prestressing force is just measured during construction using jacking device and after that, it can not be managed practically. For this reason, this study investigated the measurements of prestress using embedded sensors that can be available now with an ultimate goal to propose smart prestressed girders that can measure the prestress from the birth to the end of service life. 4 types of sensors were installed on the small prestressed girders, and the applicability and the accuracy of those sensors were tested while the prestress was applied to the girders. The results show that a center-hole type loadcell has a tendency to measure a prestressing force higher than a reference value, especially when it is loaded with an eccentricity. a EM sensor shows several advantages that has a good practical accuracy, that can be installed anyplace along the tendons.

• 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.11 no.4
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• pp.311-324
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• 2013

To ensure the reliability of computer programs used for the post-closure safety assessment in the Wolsong LILW Center, the results from MASCOT, SAFE-ROCK and GOLDSIM programs are compared with a problem for degradation. Advantages and disadvantages of each computer programs are individually analyzed. Effects on the individual dose are assessed with each computer programs. MASCOT and SAFE-ROCK showed similar results for $^{129}I$ and $^3H$. However, GOLDSIM represented different results for $^{129}I$ and $^3H$. It is analyzed further and compared with the fluxes in each barrier of the disposal system. Througout the benchmarking testing of the computer program, the limitation of computer program can be continuously found out for the mature post-closure safety of Korean radwaste disposal system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.297-305
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• 2016

An approach to assess inadvertent human intrusion into radwaste repository was proposed with the assumption that the intrusion occurs after a loss of knowledge of the hazardous nature of the disposal facility. The essential boundary conditions were derived on the basis of international recommendations, followed by an overall approach to deal with inadvertent human intrusion. The interrelation between societal factors, human intrusion scenarios, and protective measures is described to provide a concrete explanation of the approach, including the detailed procedures to set up the human intrusion scenario. The procedure for deriving protective measures is also explained with four steps, including how to derive a safety framework, general measures, potential measures, and eventual protective measures on the basis of stylized scenarios. It is expected that the approach proposed in this study will be used effectively to reduce the potential for and/or the consequences of human intrusion during the entire process of realizing a disposal facility.

• Journal of the Korean Society of Radiology
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• v.16 no.6
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• pp.741-749
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• 2022

Radiography-Testing that verify the quality of welding structures without destruction are overwhelmingly used in industries, but many safety precautions are required as radiation is used. The workers for Radiography-Testing perform the inspection by moving the Iridium-192 radiation source embedded in the transport container of the gamma-ray irradiator within or outside the facility. The general facility is completely blocked about radiation from the outside with thick concrete, but if it is difficult for worker to handle object of inspection, facilities ceiling can be opened. A general facility may be constructed using a theoretical dose evaluation method because all exterior facilities are blocked, but if the ceiling is open, it is not appropriate to evaluate radiation safety with a simple theoretical calculation method due to the skyshine effect. Therefore, in this study, the radiation safety of the facility was evaluated in the actual field through an ion chamber survey-meter and an accumulated dose-meter called as OSLD, and the actual evaluation environment was modeled and evaluated using the Monte Carlo simulation code as FLUKA. According to the direction of the irradiation, the radiation dose at the facility boundary was difficult to meet the standards set by the regulatory authority, and radiation safety could be secured through additional methods. In addition, it was confirmed that the simulation results using the Iridium-192 source were valid evaluation with the actual measured results.

• Analytical Science and Technology
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• v.35 no.6
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• pp.256-266
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• 2022

It is important to determine the concentration of long-lived radionuclides (e.g., ¹²⁹I) in nuclear waste to ensure safety when handling it. To analyze nuclides in a solid sample (e.g., concrete and soil), it is essential to effectively separate and purify the nuclides of interest in the sample. This study reports the comprehensive efforts made to validate the analytical procedure for ¹²⁹I detection in solid samples, using a high-temperature combustion furnace. ¹²⁹I volatilized from the sample collected in 0.01 M HNO₃ solution with a reducing agent (e.g., NaHSO₃) and was rapidly measured by ICP-MS. Analytical conditions, such as pyrolysis temperature and types of mobile phase gas, catalyst, and trapping solution, were optimized to obtain a high recovery rate of spiked ¹²⁹I. Finally, the optimized method was applied for the simultaneous analysis of other volatile radionuclides, such as ³H and ¹⁴C. The performance test results for the optimized method confirmed that the LSC (for ³H and ¹⁴C) and ICP-MS (for ¹²⁹I) measurements, with the separation of volatile nuclides using a high-temperature combustion furnace, were reliable.