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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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Journal DOI :
The Korean Radioactive Waste Society
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Volume & Issues
Volume 14, Issue 2 - Jun 2016
Volume 14, Issue 1 - Mar 2016
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The Effects of Impurity Composition and Concentration in Reactor Structure Material on Neutron Activation Inventory in Pressurized Water Reactor
Cha, Gil Yong ; Kim, Soon Young ; Lee, Jae Min ; Kim, Yong Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 14, issue 2, 2016, Pages 91~100
DOI : 10.7733/jnfcwt.2016.14.2.91
The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.
Adsorption Removal of Sr by Barium Impregnated 4A Zeolite (BaA) From High Radioactive Seawater Waste
Lee, Eil-Hee ; Lee, Keun-Young ; Kim, Kwang-Wook ; Kim, Ik-Soo ; Chung, Dong-Yong ; Moon, Jei-Kwon ; Choi, Jong-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 14, issue 2, 2016, Pages 101~112
DOI : 10.7733/jnfcwt.2016.14.2.101
This study investigated the removal of Sr, which was one of the high radioactive nuclides, by adsorption with Barium (Ba) impregnated 4A zeolite (BaA) from high-radioactive seawater waste (HSW). Adsorption of Sr by BaA (BaA-Sr), in the impregnated Ba concentration of above 20.2wt%, was decreased by increasing the impregnated Ba concentration, and the impregnated Ba concentration was suitable at 20.2wt%. The BaA-Sr adsorption was added to the co-precipitation of Sr with
precipitation in the adsorption of Sr by 4A (4A-Sr) within BaA. Thus, it was possible to remove Sr more than 99% at m/V (adsorbent weight/solution volume)=5 g/L for BaA and m/V >20 g/L for 4A, respectively, in the Sr concentration of less than 0.2 mg/L (actual concentration level of Sr in HSW). It shows that BaA-Sr adsorption is better than 4A-Sr adsorption in for the removal capacity of Sr per unit gram of adsorbent, and the reduction of the secondary solid waste generation (spent adsorbent etc.). Also, BaA-Sr adsorption was more excellent removal capacity of Sr in the seawater waste than distilled water. Therefore, it seems to be effective for the direct removal of Sr from HSW. On the other hand, the adsorption of Cs by BaA (BaA-Cs) was mainly performed by 4A within BaA. Accordingly, it seems to be little effect of impregnated Ba into BaA. Meanwhile, BaA-Sr adsorption kinetics could be expressed the pseudo-second order rate equation. By increasing the initial Sr concentrations and the ratios of V/m, the adsorption rate constants (
) were decreased, but the equilibrium adsorption capacities (
) were increasing. However, with increasing the temperature of solution,
was conversely increased, and
was decreased. The activation energy of BaA-Sr adsorption was 38 kJ/mol. Thus, the chemical adsorption seems to be dominant rather than physical adsorption, although it is not a chemisorption with strong bonding form.
Hydraulic Experiment for Pollutant Discharge Characteristics in a Wolseong Nuclear Power Plant Port
Yang, Byung-Mo ; Min, Byung-Il ; Park, Kihyun ; Kim, Sora ; Lee, Jung Lyul ; Suh, Kyung-Suk ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 14, issue 2, 2016, Pages 113~122
DOI : 10.7733/jnfcwt.2016.14.2.113
In this study, the dispersion process of pollutant substances in a port under wave and current environments was evaluated by a hydraulic experiment. Once the contaminants washed ashore into the port of Wolseong nuclear power plant, transport processes of pollutants were investigated by tracking the tracer according to the variations of experimental condition through a hydraulic experiment. Several hydraulic experiments were performed to analyze the pollutant discharge rate of the surface coming from the different coastal environments. From the hydraulic experiment results, the tracer concentration decreased exponentially. These results suggested that, after the tracer was transported to the open sea, a different gradient was shown under different conditions. For the case of a diluted condition, the half-life of flow rate was 2.70, 10.40, and 26.39 days for 30, 20 and 10 rpm in the left-side, respectively. The decrease of the tracer concentration under conditions of 30 rpm was much faster than that under conditions of 10 rpm. For the wave condition, the half-life of flow rate was 4.59 and 15.35 days for the right and left side of the port in a hydraulic scale prototype, respectively.
Evaluation on Radioactive Waste Disposal Amount of Kori Unit 1 Reactor Vessel Considering Cutting and Packaging Methods
Choi, Yujeong ; Lee, Seong-Cheol ; Kim, Chang-Lak ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 14, issue 2, 2016, Pages 123~134
DOI : 10.7733/jnfcwt.2016.14.2.123
Decommissioning of nuclear power plants has become a big issue in South Korea as some of the nuclear power plants in operation including Kori unit 1 and Wolsung unit 1 are getting old. Recently, Wolsung unit 1 received permission to continue operation while Kori unit 1 will shut down permanently in June 2017. With the consideration of segmentation method and disposal containers, this paper evaluated final disposal amount of radioactive waste generated from decommissioning of the reactor pressure vessel in Kori unit 1 which will be decommissioned as the first in South Korea. The evaluation results indicated that the final disposal amount from the top and bottom heads of the reactor pressure vessel with hemisphere shape decreased as they were cut in smaller more effectively than the cylindrical part of the reactor pressure vessel. It was also investigated that 200 L and 320 L radioactive waste disposal containers used in Kyung-Ju disposal facility had low payload efficiency because of loading weight limitation.
Measurement of Properties of Domestic Bentonite for a Buffer of an HLW Repository
Yoo, MalGoBalGaeBitNaLa ; Choi, Heui-ju ; Lee, Min-soo ; Lee, Seung-yeop ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 14, issue 2, 2016, Pages 135~147
DOI : 10.7733/jnfcwt.2016.14.2.135
The buffer in geological disposal system is one of the major elements to restrain the release of radionuclide and to protect the container from the inflow of groundwater. The buffer material requires long-term stability, low hydraulic conductivity, low organic content, high retardation of radionuclide, high swelling pressure, and high thermal conductivity. These requirements could be determined by the quantitative analysis results. In case of South Korea, the bentonites produced in Gyeongju area have been regarded as candidate buffer/backfill materials at KAERI (Korea Atomic Energy Research Institute) since 1997. According to the study on several physical and chemical characteristics of domestic bentonite in the same district, this is the Ca-type bentonite with about 65% of montmorillonite content. Through this study, we present the criteria for the performance evaluation items and methods when collecting new buffer/backfill materials.
Theoretical Estimation of the Impact Velocity during the PWR Spent Fuel Drop in Water Condition
Kwon, Oh Joon ; Park, Nam Gyu ; Lee, Seong Ki ; Kim, Jae Ik ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 14, issue 2, 2016, Pages 149~156
DOI : 10.7733/jnfcwt.2016.14.2.149
The spent fuel stored in the pool is vulnerable to external impacts, since the severe reactor conditions degrade the structural integrity of the fuel. Therefore an accident during shipping and handling should be considered. In an extreme case, the fuel assembly drop can be happened accidentally during handling the nuclear fuel in the spent fuel pool. The rod failure during such drop accident can be evaluated by calculating the impact force acting on the fuel assembly at the bottom of the spent fuel pool. The impact force can be evaluated with the impact velocity at the bottom of the spent fuel pool. Since fuel rods occupies most of weight and volume of a nuclear fuel assembly, the information of the rods are important to estimate the hydraulic resistance force. In this study, the hydraulic force acting on the
short rod bundle model during the drop accident is calculated, and the result is verified by comparing the numerical simulations. The methodology suggested by this study is expected to be useful for evaluating the integrity of the spent fuel.
Development of Multi-Purpose Containers for Managing LLW/VLLW from D&D
Lee, Jaesol ; Park, Jeaho ; Sung, Nakhoon ; Yang, Gehyung ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 14, issue 2, 2016, Pages 157~168
DOI : 10.7733/jnfcwt.2016.14.2.157
Radioactive waste container designs should comply with the requirements for safety (i.e., transportation, storage, disposal) and other criteria such as economics and technology. These criteria are also applicable to the future management of the large amount of LLW and VLLW to arise from decontamination and decommissioning (D&D) of nuclear power plants, which have different features compared to that of wastes from operation and maintenance (O&M). This paper proposes to develop a set of standard containers of multi-purpose usage for transportation, storage and disposal. The concepts of the containers were optimized for management of D&D wastes in consideration of national system for radioactive waste management, in particular the Gyeongju Repository and associated infrastructures. A set of prototype containers were designed and built : a soft bag for VLLW, two metallic containers for VLLW/LLW (a standard IP2 container for sea transport and ISO container for road transport). Safety analyses by simulation and tests of these designs show they are in compliance with the regulatory requirements. A further development of a container with concrete is foreseen for 2016.
Enhancement of the Life of Refractories through the Operational Experience of Plasma Torch Melter
Moon, Young Pyo ; Choi, Jang Young ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 14, issue 2, 2016, Pages 169~178
DOI : 10.7733/jnfcwt.2016.14.2.169
The properties of wastes for melting need to be considered to minimize the maintenance of refractory and to discharge the molten slags smoothly from a plasma torch melter. When the nonflammable wastes from nuclear facilities such as concrete debris, glass, sand, etc., are melted, they become acid slags with low basicity since the chemical composition has much more acid oxides than basic oxides. A molten slag does not have good characteristics of discharge and is mainly responsible for the refractory erosion due to its low liquidity. In case of a stationary plasma torch melter with a slant tapping port on the wall, a fixed amount of molten slags remains inside of tapping hole as well as the melter inside after tapping out. Nonmetallic slags keep the temperature higher than melting point of metal because metallic slags located on the bottom of melter by specific gravity difference are simultaneously melted when dual mode plasma torch operates in transferred mode. In order to minimize the refractory erosion, the compatible refractories are selected considering the temperature inside the melter and the melting behavior of slags whether to contact or noncontact with molten slags. An acidic refractory shall not be installed in adjacent to a basic refractory for the resistibility against corrosion.
Preliminary Analyses of the Deep Geoenvironmental Characteristics for the Deep Borehole Disposal of High-level Radioactive Waste in Korea
LEE, Jongyoul ; LEE, Minsoo ; CHOI, Heuijoo ; KIM, Geonyoung ; KIM, Kyungsu ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 14, issue 2, 2016, Pages 179~188
DOI : 10.7733/jnfcwt.2016.14.2.179
Spent fuels from nuclear power plants, as well as high-level radioactive waste from the recycling of spent fuels, should be safely isolated from human environment for an extremely long time. Recently, meaningful studies on the development of deep borehole radioactive waste disposal system in 3-5 km depth have been carried out in USA and some countries in Europe, due to great advance in deep borehole drilling technology. In this paper, domestic deep geoenvironmental characteristics are preliminarily investigated to analyze the applicability of deep borehole disposal technology in Korea. To do this, state-of-the art technologies in USA and some countries in Europe are reviewed, and geological and geothermal data from the deep boreholes for geothermal usage are analyzed. Based on the results on the crystalline rock depth, the geothermal gradient and the spent fuel types generated in Korea, a preliminary deep borehole concept including disposal canister and sealing system, is suggested.
Conceptual Design of a Cover System for the Degmay Uranium Tailings Site
Saidov, Vaysidin ; Kessel, David S. ; Kim, Chang-Lak ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 14, issue 2, 2016, Pages 189~200
DOI : 10.7733/jnfcwt.2016.14.2.189
The Republic of Tajikistan has ten former uranium mining sites. The total volume of all tailings is approximately 55 million tonnes, and the covered area is more than 200 hectares. The safe management of legacy uranium mining and tailing sites has become an issue of concern. Depending on the performance requirements and site-specific conditions (location in an arid, semiarid or humid region), a cover system for uranium tailings sites could be constructed using several material layers using both natural and man-made materials. The purpose of this study is to find a feasible cost-effective cover system design for the Degmay uranium tailings site which could provide a long period (100 years) of protection. The HELP computer code was used in the evaluation of potential Degmay cover system designs. As a result of this study, a cover system with 70 cm thick percolation layer, 30 cm thick drainage layer, geomembrane liner and 60 cm thick barrier soil layer is recommended because it minimizes cover thickness and would be the most cost-effective design.