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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
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 1, Issue 1 - Dec 2003
Selecting the target year
Preliminary Calculation of the Indicators of Sustainable Development for National Radioactive Waste Management Programs
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 1, issue 1, 2003, Pages 1~10
As a follow up to the Agenda 21's policy statement for safe management of radioactive waste adopted at Rio Conference held in 1992, the UN invited the LAEA to develop and implement indicators of sustainable development for the management of radioactive waste. The IAEA finalized the indicators in 2002, and is planning to calculate the member states' values of indicators in connection with operation of its Net-Enabled Waste Management Database system. In this paper, the basis for introducing the indicators into the radioactive waste management was analyzed, and calculation methodology and standard assessment procedure were simply depicted. In addition, a series of innate limitations in calculation and comparison of the indicators was analyzed. According to the proposed standard procedure, the indicators for a few major countries including Korea were calculated and compared, by use of each country's radioactive waste management framework and its practices. In addition, a series of measures increasing the values of the indicators was derived so as to enhance the sustainability of domestic radioactive waste management program.
Electrochemical Decontamination of Metallic Wastes Contaminated with Uranium Compounds
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 1, issue 1, 2003, Pages 11~23
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
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
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
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
, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion facility were performed in 1M NaNO
solution with the current density or In mA/
. it was verified that the electrochemical decontamination of the metallic wastes contaminated uranium compounds was quite successful in a NaNO
neutral salt electrolyte by reducing
radioactivities below the level of self disposal within 10 minutes regardless of the type of contaminants and the degree of contamination.
A Study on the Electrolytic Reduction Mechanism of Uranium Oxide in a LiCl-Li
O Molten Salt
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 1, issue 1, 2003, Pages 25~39
This study proposed a new electrolytic reduction technology that is based on the integration of simultaneous uranium oxide metallization and Li
O electrowinning. In this electrolytic reduction reaction, electrolytically reduced Li deposits on cathode and simultaneously reacts with uranium oxides to produce uranium metal showing more than 99% conversion. For the verification of process feasibility, the experiments to obtain basic data on the metallization of uranium oxide, investigation of reaction mechanism, the characteristics of closed recycle of Li
O and mass transfer were carried out. This evolutionary electrolytic reduction technology would give benefits over the conventional Li-reduction process improving economic viability such as: avoidance of handling of chemically active Li-LiCl molten salt increase of metallization yield, and simplification of process.
Long-term leach rates of simulated borosilicate waste glasses under a repository condition
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 1, issue 1, 2003, Pages 41~46
To understand the long-term leach behavior of a borosilicate Waste glass in a repository, the leaching experiment with three kinds of simulated borosilicate waste glasses has been carried out since the middle of 1997. The five years results indicate that a boron would be applied as an indicator of a long-term leaching of their borosilicate waste glasses and that their long-term leach rates have a tendency to be close to about 0.03g/
-day even though their compositions and their ratios of the surface area to the volume of leachate are different.
Batch Scale Mock-up Test for the Electrochemical Reduction of Spent Oxide Fuel
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 1, issue 1, 2003, Pages 47~53
An electrochemical reduction technology which can reduce the decay heat, volume, and radioactivity of spent fuel by a factor of quarter through converting oxide type spent fuel to a metallic form in a molten salt was developed and tests in a scale of g (3- 40g
batch) have been carried out by Korea Atomic Energy Research Institute. In this research, the reaction apparatus in a scale of 5kg
batch was designed and manufactured for the mock-up test to obtain design data of the apparatus which will be used for the hot test in a scale of 20kg
batch. The electrochemical reduction behavior of
was analyzed regarding the operational factors and fresh
powder was metallized with a more than 99% yield verifying the process validity of electrochemical reduction process in a kg scale.
Development of the Maintenance Process Based on Graphic Simulation for the Parts of the Equipment at the outside of the MSM′s Workspace in a Hot Cell
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 1, issue 1, 2003, Pages 55~64
In this study, the maintenance process by the servo manipulator has been developed for the parts of the equipment, which we unable to reach out by the Master-Slave Manipulator(MSM) in a hot cell. To do this, a virtual mock-up is implemented using the iか prototyping technology. Using this mock-up, the workspace of the manipulators in the hot cell and the operator's view through the wall-mounted lead glass have been analyzed. In addition, the path planning of the servo manipulator using the collision detection function of the virtual mock-up has been established. From these, the maintenance process for the parts of the equipment, which are located at the outside of the MSM's workspace using the servo manipulator has been proposed and verified through the graphic simulation. It is revealed that the proposed remote maintenance process of the equipment can effectively be used in the real hot cell operation. It is also believed that the implemented virtual mock-up of the hot cell can effectively be applied in analyzing the various hot cell operation and enhancing the reliability and safety in a hot cell remote handling for the spent fuel management.
Removal Efficiency of Organic Iodide on Silver Ion-Exchanged Yeolite and TEDA-AC at High Temperature Process
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 1, issue 1, 2003, Pages 65~72
Adsorption and desorption characteristics of methyl iodide at high temperature conditions up to 25
by TEDA-impregnated activated carbon and silver-ion exchanged zeolite(AgX-10), which are used for radioiodine retention in nuclear facility, were experimentally evaluated. In the range of temperature from 3
, the adsorption capacity of base activated carbon decreased sharply with increasing temperature but that of TEDA-impregnated activated carbon showed higher value even at high temperature ranges. Especially, the residual amount of methyl iodide after desorption on TEDA-AC represented 30% lower value than that on AgX-10. However, it can be used as an adsorbent for the removal of methyl iodide up to 15
if it is preventing explosion by Ignition. The breakthrough curves of methyl iodide in the fixed bed packed with AgX-10 uP to 40
were compared upon the effects of bed temperatures, bed depth and input concentration of methyl iodide. Removal mechanism of methyl iodide on AgX-10 was proposed, based on the analysis of by-product gas generated from adsorption reaction.
A Study on the Removal Method of Radioactive Corrosion Product using its Magnetic Property
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 1, issue 1, 2003, Pages 73~79
In a pressurized water reactor, radioactive corrosion products (CRUD) in primary coolant system are one of the major sources for the occupational radiation exposure of the personnel in a nuclear power plant. Through the recent trend of long term fuel cycle in a nuclear power Plant, radioactive corrosion products deposited in reactor core for a long time are also the cause of Axial Offset Anomaly (AOA) having m effect on reactor power by the hideout of boron. CRUD consist primarily of magnetite, nickel ferrite, cobalt ferrite, and so on. They have the characteristic of strong magnetism. Therefore it is performed the conceptual design to develop the filter which removes the CRUD by magnetic field that is generated by an arrangement of permanent and electric magnets. Contrary to the conventional filter, the proposed filter does not interrupt the fluid flow, so there is little pressure drop and it can be used continuously. It is expected to be applied as one of the technologies for removal of the CRUB.
Prediction of the Dynamic Adsorption Behaviors of the Uranium and Cobalt Ions in a Fixed Bed by Surface Modified Activated Carbon
Geun-IL Park ; Jung-Won Lee ; Kee-Chan Song ; In-Tae Kim ; Kwang-Wook Kim ; Myung-Seung Yang ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 1, issue 1, 2003, Pages 81~92
In order to predict the dynamic behaviors of uranium and cobalt in a fixed bed at various influent pH values of liquid waste, the adsorption system is regarded as a multi-component adsorption between each ionic species in the solution. Langmuir isotherm parameters of each species were extracted by incorporating equilibrium data with the solution chemistry of the uranium and cobalt using IAST. Prediction results were in good agreement with the experimental data, except for a high concentration and pH. Although there was some limitations in predicting the cobalt adsorption, this method may be useful in analyzing a complex adsorption system where various kinds of ionic species exist in a solution.
Comparative Evaluation of Various Standard Methods in Leaching Test of Radioactive Waste Form
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 1, issue 1, 2003, Pages 93~103
IAEA, FT-04-020, and ANS 16.1, standard leaching test methods, were evaluated comparatively with their test results. Leaching index of
Cs by ANS 16.1 method for waste forms of paraffin and cement were above 6.0. Their leaching behavior were depending on the type of matrix and leachant. Leachability of
Co for cement waste form was higher in simulated seawater than do-mineralized water, and was higher in de-mineralized water for paraffin waste form. leachability of
Co was contrary to
Cs. Cumulative fraction leached of
Co was higher in order or IAEA ＞ ANS ＞ FT in a cement waste form.