• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.1
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• pp.55-64
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• 2023

There is a gap in our understanding of the behavior of fused and molten fuel salts containing unavoidable contamination, such as those due to fabrication, handling, or storage. Therefore, this work used calorimetry to investigate the change in liquidus temperature of PuCl₃, having an unknown purity and that had been in storage for several decades. Further research was performed by additions of NaCl, making several compositions within the binary system, and summarizing the resulting changes, if any, to the phase diagram. The melting temperature of the PuCl₃ was determined to be 746.5℃, approximately 20℃ lower than literature reported values, most likely due to an excess of Pu metal in the PuCl₃ either due to the presence of metallic plutonium remaining from incomplete chlorination or due to the solubility of Pu in PuCl₃. From the melting temperature, it was determined that the PuCl₃ contained between 5.9 to 6.2mol% Pu metal. Analysis of the NaCl-PuCl₃ samples showed that using the Pu rich PuCl₃ resulted in significant changes to the NaCl-PuCl₃ phase diagram. Most notably an unreported phase transition occurring at approximately 406℃ and a new eutectic composition of 52.7mol% NaCl-38.7mol% PuCl₃-2.5mol% Pu which melted at 449.3℃. Additionally, an increase in the liquidus temperatures was seen for NaCl rich compositions while lower liquidus temperatures were seen for PuCl₃ rich compositions. It can therefore be concluded that changes will occur in the NaCl-PuCl₃ binary system when using PuCl₃ with excess Pu metal. However, melting temperature analysis can provide valuable insight into the composition of the PuCl₃ and therefore the NaCl-PuCl₃ system.

• Journal of Radiation Industry
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• v.17 no.3
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• pp.283-292
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• 2023

Spent nuclear fuel corresponds to high-level radioactive waste that has high decay heat and radioactivity. Accordingly, Spent nuclear fuel withdrawn from the reactor core is primarily stored and managed in a spent nuclear fuel pool in the nuclear power plant to reduce decay heat and radioactivity. In Korea, most nuclear power plant store all spent nuclear fuel in a spent nuclear fuel pool. For wet storage, there are no defense in depth different with reactor core. The study related to spent nuclear fuel pool accident should be carried out to ensure safety. Therefore, it is necessary to analyze previous study cases related to safety of spent nuclear fuel pool and accident cases to build foundational knowledge. The Objective of this study is to analyze study cases of safety assessment and cases for spent nuclear fuel pool accident. For analyzing study cases of safety assessment, possible phenomena when spent nuclear fuel pool accident occurring identified, Subsequently, study cases for safety assessment about each phenomena were investigated, and materials & methods and results for each study are analyzed. For analyzing cases for spent nuclear fuel pool accident, we analyzed accident cases caused by loss of cooling and loss of coolant in spent nuclear fuel pool. Subsequently, causes and change of water level and temperature by each accident case are analyzed. As a result of the analysis on study cases of spent nuclear fuel pool accident, the results of the study conducted by each research institute were vary depending on the computer code, materials & methods of experiment and major assumptions used in the study. As a result of analyzing cases for spent nuclear fuel pool accident, it was found that accident cases for loss of cooling is more than cases for loss of coolant accident. Even though the types of accident in spent nuclear fuel pool were similar, the specific causes were different by each accident case. All the accident cases analyzed did not lead to severe accidents, such as nuclear fuel being exposed to the air. The result of this study will be used as fundamental data for study on spent nuclear fuel pool accident that will be conducted in the future.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2022.04a
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• pp.166-167
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• 2022

The International Energy Agency(IEA) recommends that intergovernmental agreements reduce CO2 emissions by 2050 to about 50% in 2005 in its report. To realize these demands, it is suggested to actively utilize energy efficiency improvement technology, renewable energy, nuclear power, carbon dioxide capture & storage technology (CCS). In the field of building materials and cement, mineral carbonization technology is widely used. Inorganic by-products applicable to greenhouse gas storage include waste concrete, slag, coal ash, and gypsum. If the Mineral Carbonation Act is used, it is expected that about 12 million tons of greenhouse gases can be immobilized every year. Greenhouse gas immobilization using cement hydrate can be immobilized by injecting carbon dioxide into the hydrated products C-S-H, and Ca(OH)2. In the case of immobilization through concrete carbonization, a carbon dioxide promotion test is used, which is often different from the actual carbon dioxide carbonization reaction. If the external carbon dioxide concentration is abnormally higher than the reality, it is thought that it will be different from the actual reaction. In this study, the carbonation phenomenon according to the concentration and identification of the carbon dioxide reaction mechanism of cement hydrate was to be considered.

• Journal of Radiation Protection and Research
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• v.45 no.2
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• pp.81-87
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• 2020

Background: Kori unit #1 is permanently shut down after a 40-year lifetime. The Nuclear Safety and Security Commission recommends establishing initial decommissioning plans for all nuclear and radwaste treatment facilities. Therefore, the Korea Atomic Energy Research Institute (KAERI) must establish an initial and final decommissioning plan for radwaste-treatment facilities. Radiation safety assessment, which constitutes one chapter of the decommissioning plan, is important for establishing a decommissioning schedule, a strategy, and cost. It is also a critical issue for the government and public to understand. Materials and Methods: This study provides a method for assessing external radiation dose to workers during decommissioning. An external dose is calculated following each exposure scenario, decommissioning strategy, and working schedule. In this study, exposure dose is evaluated using the deterministic method. Physical characterization of the facility is obtained by both direct measurement and analysis of the drawings, and radiological characterization is analyzed using the annual report of KAERI, which measures the ambient dose every month. Results and Discussion: External doses are calculated at each stage of a decommissioning strategy and found to increase with each successive stage. The maximum external dose was evaluated to be 397.06 man-mSv when working in liquid-waste storage. To satisfy the regulations, working period and manpower must be managed. In this study, average and cumulative exposure doses were calculated for three cases, and the average exposure dose was found to be about 17 mSv/yr in all the cases. Conclusion: For the three cases presented, the average exposure dose is well below the annual maximum effective dose restriction imposed by the international and domestic regulations. Working period and manpower greatly affect the cost and entire decommissioning plan; hence, the chosen option must take account of these factors with due consideration of worker safety.

• Proceedings of the Korean Geotechical Society Conference
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• 2000.11a
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• pp.705-712
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• 2000

The fluid generally flows through fractures in crystalline rocks where most of underground storage facilities are constructed because of their low hydraulic conductivities. The fractured rock is better to be conceptualized with a discrete fracture concept, rather continuum approach. In the aspect of fluid flow in underground, the simultaneous flow of groundwater and gas should be considered in the cases of generation and leakage of gas in nuclear waste disposal facilities, air sparging process and soil vapor extraction for eliminating contaminants in soil or rock pore, and pneumatic fracturing for the improvement of permeability of rock mass. For the purpose of appropriate analysis of groundwater-gas flow, this study presents an unsteady-state multi-phase FEM fracture network simulator. Numerical simulation has been also conducted to investigate the hydraulic head distribution and air tightness around Ulsan LPG storage cavern. The recorded hydraulic head at the observation well Y was -5 to -10 m. From the results obtained by the developed model, it shows that the discrete fracture model yielded hydraulic head of -10 m, whereas great discrepancy with the field data was observed in the case of equivalent continuum modeling. The air tightness of individual fractures around cavern was examined according to two different operating pressures and as a result, only several numbers of fractures neighboring the cavern did not satisfy the criteria of air tightness at 882 kPa of cavern pressure. In the meantime, when operating pressure is 710.5 kPa, the most areas did not satisfy air tightness criteria. Finally, in the case of gas leaking from cavern to the surrounding rocks, the resulted hydraulic head and flowing pattern was changed and, therefore, gas was leaked out from the cavern ceiling and groundwater was flowed into the cavern through the walls.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4204-4212
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• 2023

Background: Storage reservoirs of radioactive waste could be the source of atmospheric pollution due to the efflux of aqueous aerosol from their water areas. The main mechanism of formation of aqueous aerosols is the collapse of gas bubbles at the water surface. In this paper, we discuss the potential influence of biological factors on gas ebullition in the water areas of the radioactively contaminated industrial reservoirs R-9 (Lake Karachay) and R-4 (Metlinsky pond) of the Mayak PA. The emission of the released non-dissolved gases captured with gas traps in reservoir R-9 was (88-290) ml/m² per day (2015) and in reservoir R-4 (270-460) ml/m² per day (2016). The analysis of gas composition in reservoir R-4 (60% methane, 35% nitrogen, 2.4% oxygen, 1.5% carbon dioxide) confirms their biological origin. It is associated with the processes of organic matter destruction in bottom sediments. The major source of organic matter in bottom sediments is the dying phytoplankton developing in these reservoirs. Conclusion: The obtained results form the basis to set a task to quantify the relationship between the phytoplankton development, gases ebullition and radioactive atmosphere contamination.

• Journal of the Korean Professional Engineers Association
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• v.37 no.6
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• pp.29-36
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• 2004

Human being have destroyed environment through rapid industrialization, so that the world is increasingly faced with golbal environmental problem such as destruction of ozone layer, global warming., acid rain, decertification and we are heading toward impending disaster of the earth. Especially, recent environmental pollution of China from rapid economic development threatens neighboring Korea and Japan. Regional development are necessary for promotion of national prosperity, but theses regional developments must have brought about environmental and preservation. Actually there are so many conflict cases; Saemankum reclamation project, construction of nuclear waste storage facility, dam construction of China etc. Here, harmony and balance between development and preservation are needed to solve these great crux. Under theses condition, the concept of ESSD(Environmentally Sound and Sustainable Development) emerged as a global issue. And staring point to solve present environmental problems is to perceive the truth that we have inherited only one earth and hence owe a debt to preserve its environment for the benefit of the future generations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.239-245
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• 2010

There are 11,811 LWR spent fuels stored at reactor sites, as of 2009. Source terms based on reference spent fuel which represents entire spent fuels with bounding values in the aspect of source term has been applied to a design of nuclear installations, instead of those which are generated by weighting respective source term for each spent fuel. Simplified regression models to estimate total decay heat, radioactivity, and ingestion hazard index for spent fuel from Westinghouse-type reactors were developed in this study, because it can be used as a fundamental model for weighting source term for respective spent fuel to exclude conservativeness in source terms. It was found that the estimated source terms agreed with calculated value from ORIGEN-ARP within 5%. It was also found that the conservativeness could be excluded if the weight source terms were used as reference source term in the design. Therefore, it is expected that the developed regression model could be widely used in the conceptual design process of nuclear facilities related with storage and disposal of spent nuclear fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.65-78
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• 2007

The pollutants (Rn, CH, CO, HS, radioactive gas from radiolysis) were generated from the process of construction and operation of underground repository, and after disposal of low-intermediate radioactive waste inside there must be controlled by a ventilation system to distribute them in area where enough air is supported. Therefore, a suitable technical approach is needed especially at an underground repository that is equipped with many entry tunnels, storage tunnels, exhaust-blowing tunnels, and vertical shafts in complicated network form. For the technical approach of such a ventilation system, WIPP (Waste Isolation Pilot Plant) in U. S and SFR (Slutforvar for Reaktorafall) low-intermediate radioactive waste repository in Sweden were selected as the models, for calculating the required air quantity, organizing a ventilation network considering cross section, length, surface roughness of the air passage, and describing a calculation of resistance of each circuit. Based on these procedures, a best suited ventilation system was completed with designing proper capacity of fans and operating plan of vertical shafts. As a result of comparing the two repositories based on the geometry dimensions and ventilation facility equipment operation, more parallel circuit as in WIPP, brought decrease in resistance for entire system leading to reduce of operating costs, and the larger cross-sectional area of the SFR, the greater the percentage of disposal capacity. Accordingly, the mixture of parallel circuit of WIPP repository for reducing resistance and SFR repository formation for enlargement of disposal capacity would be the most rational and efficient ventilation system.