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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)
Journal Basic Information
Journal DOI :
The Korean Radioactive Waste Society
Editor in Chief :
Volume & Issues
Volume 8, Issue 4 - Dec 2010
Volume 8, Issue 3 - Sep 2010
Volume 8, Issue 2 - Jun 2010
Volume 8, Issue 1 - Mar 2010
Selecting the target year
Study on the Synthesis Method of Simulated CRUD for Chemical Decontamination in NPPs
Kang, Duk-Won ; Kim, Jin-Kil ; Kim, Kyeong-Sook ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 2, 2010, Pages 91~97
As nuclear power plants are getting older, interests on a decontaminating process are increasingly attracting more attention. Chemical decontamination is crucial to lower the production of radioactive waste and radiation dose rate. Prior to this, oxidizers and detergents for target material should be chosen so as to decontaminate major systems and components of a nuclear power plant chemically. In order to decontaminate it properly, it is crucial to have information about the chemical composition and crystalline structure of CRUD, analyzing its samples from the target or the decontamination system with components. However, there is no program which enables the extraction of samples directly from the object or the decontamination system with components carrying genuine radioactivity. Therefore, it is limited to samples from corrosion products carrying partial radioactivity as a resource. The composition of CRUD varies considerably depending on refueling cycle because it is closely related to the constituent of basic material. After settling a target, it is crucial to analyze and obtain analytical information about CRUD as a decontamination target. In this paper, various technologies for manufacturing simulated CRUD are introduced as alternatives to unattained samples. A metal oxide or metal hydroxide was used to synthesize simulated cruds having chemical compositions and crystalline stricture similar to the actual one by 12 different methods. CRUD 4(metal oxides in the autoclave vessel) and CRUD 10(metal oxides in a crucible after hydrazing pretreatment)were chosen as the best method for Type 1 and Type 2.respectively. As these CRUD can be synthesized easily without using any specialized equipment or reagents in a short time and in large quantities, they are expected to stimulate the development of decontaminating agents and processes.
A Study on Wasteform Properties of Spent Salt Treated with Zeolite and SAP
Kim, Hwan-Young ; Park, Hwan-Seo ; Kang, Kweon-Ho ; Ahn, Byung-Gil ; Kim, In-Tae ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 2, 2010, Pages 99~105
This paper investigated the characteristics of wasteform containing a spent zeolite used as a separating agent of FPs for recycling LiCl waste which would be generated from pyrochemical process of spent PWR fuel. In this study, a conventional borosilicate and Ca-rich glass were used as a consolidating agent for spent zeolite and it's mixing ratio was changed in the range,
. The leach rates of Cs and Sr had about
, respectively. The leach resistance of Cs increased with amount of SAP and it showed about 10 times higher in the Ca-rich glass wasteform than in the conventional borosilciate glass wasteform. The compressive strength of wasteform was affected with the amount of glass. Thermal expansion rate of containing 30wt% glass has relatively lower than others. Also, the melting temperature was little changed in given mixing ratio of glass.
A Study of Adsorption Behaviour of Humic Acid and Americium on the Kaolinite
Lee, Myung-Ho ; Lee, Kyu-Whan ; Park, Kyung-Kyun ; Jung, Euo-Chang ; Song, Kyu-Seok ; Shin, Hyun-Sang ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 2, 2010, Pages 107~113
In this study, the adsorption reactions in the binary component system such as kaolinite-humic acid, kaolinite-americium and humic acid-americium were investigated. After performing the basic physico-chemical properties of the kaolinite, the adsorption reactions of the humic acid on the kaolinite were carried out with varying concentration of humic acid and ion strength, and pH. With increasing HA concentration and pH, the sorption of HA onto KA decreased, while the sorption of HA onto KA increased with increasing ionic stre ngth. Also, with varying pH, the adsorption reactions of the americium-kaolinite and americium-humic acid were studied. In the acid and neutral region, Am easily adsorbed on the HA, while the sorption of Am on the HA in the alkali region decreased because of electrostatic repulsion. The results from these studies make it possible to understand the characteristics of adsorption behaviour of the americium by the humic acid in the water environment.
Precipitation behaviors of Cs and Re(/Tc) by NaTPB and TPPCl from a simulated fission products-
Lee, Eil-Hee ; Lim, Jae-Gwan ; Chung, Dong-Yong ; Yang, Han-Beum ; Kim, Kwang-Wook ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 2, 2010, Pages 115~122
In this study, the removal of Cs and Tc from a simulated fission products (FP) solution which were co-dissolved with U during the oxidative-dissolution of spent fuel in a mixed carbonate solution of
was investigated by using a selective precipitation method. As Cs and Tc might cause an unstable behavior due to the high decay heat emission of Cs as well as the fast migration of Tc when disposed of underground, it is one of the important issues to removal them in views of the increase of disposal safety. The precipitation of Cs and Re (as a surrogate for Tc) was examined by introducing sodium tetraphenylborate (NaTPB) and tetraphenylphosponium chloride (TPPCl), respectively. Precipitation of Cs by NaTPB and that of Re by TPPCl were completed within 5 minutes. Their precipitation rates were not influenced so much by the temperature and stirring speed even if they were increased by up to
and 1,000 rpm. However, the pH of the solution was found to have a great influence on the precipitation with NaTPB and TPPCl. Since Mo tends to co-precipitate with Re at a lower pH, especially, it was effective that a selective precipitation of Re by TPPCl was carried out at pH of above 9 without co-precipitation of Mo and Re. Over 99% of Cs was precipitated when the ratio of [NaTPB]/[Cs]>1 and more than 99% of Re, likewise, was precipitated when the ratio of [TPPCl]/[Re]>1.
Statistical Approach for Determination of Compliance with Clearance Criteria Based upon Types of Radionuclide Distributions in a Very Low-Level Radioactive Waste
Cheong, Jae-Hak ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 2, 2010, Pages 123~133
A statistical evaluation methodology was developed to determine the compliance of candidate waste stream with clearance criteria based upon distribution of radionuclide in a waste stream at a certain confidence level. For the cases where any information on the radionuclide distribution is not available, the relation between arithmetic mean of radioactivity concentration and its acceptable maximum standard deviation was demonstrated by applying widely-known Markov Inequality and One-side Chebyshev Inequality. The relations between arithmetic mean and its acceptable maximum standard deviation were newly derived for normally or lognormally distributed radionuclide in a waste stream, using probability density function, cumulative density function, and other statistical relations. The evaluation methodology was tested for a representative case at 95% of confidence level and 100 Bq/g of clearance level of radioactivity concentration, and then the acceptable range of standard deviation at a given arithmetic mean was quantitatively shown and compared, by varying the type of radionuclide distribution. Furthermore, it was statistically demonstrated that the allowable range of clearance can be expanded, even at the same confidence level, if information on the radionuclide distribution is available.
Empirical model to estimate the thermal conductivity of granite with various water contents
Cho, Won-Jin ; Kwon, Sang-Ki ; Lee, Jae-Owan ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 2, 2010, Pages 135~142
To obtain the input data for the design and long-term performance assessment of a high-level waste repository, the thermal conductivities of several granite rocks which were taken from the rock cores from the declined borehole were measured. The thermal conductivities of granite were measured under the different conditions of water content to investigate the effects of the water content on the thermal conductivity. A simple empirical correlation was proposed to predict the thermal conductivity of granite as a function of effective porosity and water content which can be measured with relative ease while neglecting the possible effects of mineralogy, structure and anisotropy. The correlation could predict the thermal conductivity of granite with the effective porosity below 2.7% from the KURT site with an estimated error below 10%.
Feasibility Study on the Vitrification of Concentrated Boric Acid Waste
Cho, Hyun-Je ; Kim, Deuk-Man ; Park, Jong-Kil ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 2, 2010, Pages 143~150
Vitrification technology has been gradually recognized as one of effective solidification methods for concentrated boric acid wastes generated in PWR. Vitrification for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. A feasibility study for the vitrification of concentrated boric acid wastes has been performed with developing the pre-treatment methods of powdered wastes, glass compositions using glass formulation and demonstration test. The pre-treatment method is pelletizing the powder type for stable feeding within cold crucible melter. The glass compositions should be developed considering molten glass are related with wastes reduction. High contents of sodium and boron within borate wastes give influence to waste loading. A variety of factors obtained from the demonstration test are reviewed, which is wastes feeding rate, off-gas characteristics on stack and glass characteristics of final products such as durability for implementing the wastes disposal requirement. The aim of this paper is to present the feasibility of vitrification and review the solidification method for concentrated boric acid wastes and obtain the physicochemical characteristics of solidified glass.
Deposition Velocity of Iodine Vapor (
) for Radish Plants and Its Root-Translocation Factor : Results of Experimental Exposures
Choi, Yong-Ho ; Lim, Kwang-Muk ; Jun, In ; Park, Doo-Won ; Keum, Dong-Kwon ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 2, 2010, Pages 151~158
In order to measure the deposition velocity of
vapor for radish plants and its translocation factor for their roots, radish plants were exposed to
vapor for 80 min. at different growth stages between 29 and 53 d after sowing. The exposure was performed in a transparent chamber during the morning time. Deposition velocities (
) were on the whole in the range of
showing an increasing tendency with an increase in the biomass density. The results showed some agreement with existing reports that a higher relative humidity would lead to a higher deposition velocity. The acquired deposition velocities were lower than by factors of several tens than some field measurements probably due to a very low wind speed (about
) in the chamber. Translocation factors (ratio of the total iodine in the roots at harvest to the total plant deposition), estimated in a more or less conservative way, were
for an exposure at 29 d after sowing and
for an exposure at 53 d after sowing. In using the present experimental data, meteorological conditions and chemical and physical forms of iodine need to be carefully considered.
Groundwater Flow Modeling in a Block-Scale Fractured Rocks considering the Fractured Zones
Ko, Nak-Youl ; Ji, Sung-Hoon ; Koh, Yong-Kwon ; Choi, Jon-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 2, 2010, Pages 159~166
The block-scale groundwater flow system at Olkiluoto site in Finland was simulated. The heterogeneous and anisotropic hydraulic conductivity field for the domain was constructed from the discrete fracture network, which considered only the fractured zones identified in the deep boreholes installed in the study site. The groundwater flow model was calibrated by adjusting the recharge rate and the transmissivities of the fractured zones to fit the calculated hydraulic heads and into- and out-flow rates in the observation intervals of the boreholes with the observed ones. In the calibrated model, the calculated flow rates at some intervals were not in accordance with the observed ones although the calculated hydraulic heads fit well with the observed ones, which revealed that the number of the conduits for groundwater flow is insufficient in the conceptual model for groundwater flow modeling. Therefore, it was recommended that the potential local conduits such as background fractures should be added to the present conceptual model.
Some notes on the Timing of Geological Disposal of CANDU Spent Fuels
Choi, Heui-Joo ; Kook, Dong-Hak ; Choi, Jong-Won ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 2, 2010, Pages 167~172
CANDU spent fuel is to be disposed of at repository finally rather than recycled because of its low fissile nuclide concentration. But the difficult situation of finding a repository site can not help introducing a interim storage in the short term. It is required to find an optimum timing of geological disposal of CANDU spent fuels related to the interim storage operation period. The major factors for determining the disposal starting time are considered as safety, economics, and public acceptance. Safety factor is compared in terms of the decay heat and non-proliferation. Economics factor is compared from the point of the operation cost, and public acceptance factor is reviewed from the point of retrievability and inter-generation ethics. This paper recommended the best solution for the disposal starting time by analyzing the above factors. It is concluded that the optimum timing for the CANDU spent fuel disposal is around 2041 and that the sooner disposal time, the better from the point of technical and safety aspects.