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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 7, Issue 4 - Dec 2009
Volume 7, Issue 3 - Sep 2009
Volume 7, Issue 2 - Jun 2009
Volume 7, Issue 1 - Mar 2009
Selecting the target year
Thermal Release of LiCl Waste Salt from Pyroprocessing
Kim, Jeong-Guk ; Kim, Kwang-Rag ; Kim, In-Tae ; Ahn, Do-Hee ; Lee, Han-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 7, issue 2, 2009, Pages 73~78
The decay heat of Cs and Sr contained in a LiCl waste salt, generated from an electrolytic reduction process in pyroprocessing of spent nuclear fuel, has been calculated. The calculation has been carried out under some assumptions that most of the LiCl waste is purified and recycled to main process, and the residual is fabricated to make a waste form. As a result, the decay heat from daughter nuclides such as Ba and Y seems to be maximum 4.6 times higher than that from their parent nuclides such as Cs and Sr. The thermal release from Cs and Sr in the LiCl waste is the maximum around the first one month, so an cooling system operation for some time at the beginning would be suggested to control a rapid increase in the temperature of the LiCl waste salt.
Separation of Radionuclide from Dismantled Concrete Waste
Min, Byung-Youn ; Park, Jung-Woo ; Choi, Wang-Kyu ; Lee, Kune-Woo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 7, issue 2, 2009, Pages 79~86
Concrete materials in nuclear facilities may become contaminated or activated by various radionuclides through different mechanism. Decommissioning and dismantling of these facilities produce considerable quantities such as concrete structure, rubble. In this paper, the characteristics distribution of the radionuclide have been investigated for the effects of the heating and grinding test for aggregate size such as gravel, sand and paste from decommissioning of the TRIGA MARK II research reactor and uranium conversion plant. The experimental results showed that most of the radionuclide could be removed from the gravel, sand aggregate and concentrated into a paste. Especially, we found that the heating temperature played an important role in separating the radionuclide from the concrete waste. Contamination of concrete is mainly concentrated in the porous paste and not in the dense aggregate such as the gravel and sand. The volume reduction rate could be achieved about 80% of activated concrete waste and about 75% of dismantled concrete waste generated from UCP.
Development of an Organic Scintillator Sensor for Radiation Dosimetry using Transparent Epoxy Resin and Optical Fiber
Park, Chan-Hee ; Seo, Bum-Kyoung ; Lee, Dong-Gyu ; Lee, Kune-Woo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 7, issue 2, 2009, Pages 87~92
Remote detecting system for a radiation contamination using a plastic scintillator and an optical fiber was developed. Using a commercially available silica optical fiber and a plastic scintillator, we tested then for a real possibility as a remote monitoring detector. Also, a plastic scintillator was developed by itself, and evaluated as a radiation sensor. The plastic scintillator was made of epoxy resin, a hardener and an organic scintillation material. The mixture rate of the epoxy resin, hardener and organic scintillator was fixed by using their emission spectrum, transmittance, intensity etc. In this study, in order to decrease the light loss of an incomplete connection between an optical fiber and a scintillator, the optical fiber was inserted into the scintillator during the fabrication process. The senor used a plastic optical fiber and was estimated for its detection efficiency by an optic fiber's geometric factor.
The Characteristics of an Oxidative Dissolution of Simulated Fission Product Oxides in
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 7, issue 2, 2009, Pages 93~100
This study has been carried out to look into the characteristics of an oxidative-dissolution of fission products (FP) co-dissolved with uranium (U) in a
carbonate solution. Simulated FP-oxides which contained 12 components have been added to the solution to examine their dissolution characteristics. It is found that
is an effective oxidant to minimize the oxidative-dissolution of FP. In the 0.5 M
solution, some elements such as Re, Te, Cs and Mo seem to be dissolved together with U, while 98
2% for Re and Te, 94
2% for Cs, and 29
2 % for Mo are dissolved for 2 hours. It is revealed that dissolution rates of Re, Te and Cs are high (completely dissolved within 10
20 minutes) due to their high solubility in the
solution regardless of the addition of
, and independent of the concentrations of
. However, the dissolution ratio of Mo seems to be slightly increased with time and about 33 % for 4 hours, indicating a very slow dissolution rate and also independent of the
concentration. It is found that the most important factor for the oxidative-dissolution of FP is the pH of the solution and an effective dissolution is achieved at a pH between 9
10 in order to minimize the dissolution of FP.
Determination of Water Content in Compacted Bentonite Using a Hygrometer and Its Application
Lee, Jae-Owan ; Cho, Won-Jin ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 7, issue 2, 2009, Pages 101~107
Investigation of resaturation and thermal-hydro-mechanical behavior for the buffer of a repository requires measuring the water content of compacted bentonite. This study investigated the relative humidity of compacted bentonites using a humidity sensor (Vaisala HMT 334) applicable under high temperature and pressure, and then conducted a multi-regression analysis based on the measured results to determine relationships among the water content, relative humidity, and temperature. The relationships for the compacted bentonites with the dry densities of 1,500
were expressed as
, respectively. These were then used to interpret the resaturation of bentonite blocks in the KENTEX test.
A Sensitive Detection of Actinide Species in Solutions Using a Capillary Cell
Cho, Hye-Ryun ; Park, Kyuong-Kyun ; Jung, Euo-Chang ; Song, Kyu-Seok ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 7, issue 2, 2009, Pages 109~114
Absorption spectra for a quantitative analysis of actinide elements such as U(VI) and Pu(V) were measured by using a liquid waveguide capillary cell (LWCC) which has an optical path length of 1.0 meter. In order to investigate radioactive elements, a LWCC is installed in a glove box and is coupled to a spectrophotometer with optical fibers. Limits of detection (LOD) for the system were determined as 0.74 and 0.35 M with molar absorption coefficients of 8.14
0.07 (414 nm) and 17.00
0.16 (569 nm)
for U(VI) and Pu(V) ions, respectively. The measured LOD values are about 30 times more sensitive when compared to those achievable by using a conventional quartz cell with an optical path length of 1.0 cm. As an application with an enhanced sensitivity, a quantitative analysis for micromolar concentrations of Pu(V) has been performed to decrease the uncertainty in the formation constant of the Pu(VI)-OH complex.
Feasibility Study of Cryogenic Cutting Technology by Using a Computer Simulation and Manufacture of Main Components for Cryogenic Cutting System
Kim, Sung-Kyun ; Lee, Dong-Gyu ; Lee, Kune-Woo ; Song, Oh-Seop ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 7, issue 2, 2009, Pages 115~124
Cryogenic cutting technology is one of the most suitable technologies for dismantling nuclear facilities due to the fact that a secondary waste is not generated during the cutting process. In this paper, the feasibility of cryogenic cutting technology was investigated by using a computer simulation. In the computer simulation, a hybrid method combined with the SPH (smoothed particle hydrodynamics) method and the FE (finite element) method was used. And also, a penetration depth equation, for the design of the cryogenic cutting system, was used and the design variables and operation conditions to cut a 10 mm thickness for steel were determined. Finally, the main components of the cryogenic cutting system were manufactures on the basis of the obtained design variables and operation conditions.
Performance Evaluation of Stirrers for Preventing Dendrite Growth on Liquid Cathode
Kim, Si-Hyung ; Yoon, Dal-Seong ; You, Young-Jae ; Paek, Seung-Woo ; Shim, Joon-Bo ; Ahn, Do-Hee ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 7, issue 2, 2009, Pages 125~131
An electrolytic system (zinc anode-gallium cathode) was setup to evaluate the performance of several stirrers prepared for this study, where stirrers have been used to prevent uranium from forming dendrite on the cathode in pyrochemical process. In the case of no-stirring condition, zinc dendrites began to grow on the gallium surface in 1 hour and some dendrite grew out of the cathode crucible around 6 hours. When a rectangular stirrer or a tilt stirrer was rotated, at 40
150 rpm, to mix the liquid gallium cathode, dendritic growth of zinc metal was prevented irrespective of revolution speed, but some of the deposits overflowed out of the cathode crucible owing to the large centrifugal forces at 150 rpm. The harrow stirrer did not nearly retard the dendrite growth at 40 rpm, but the dendrite growth was retarded at higher than 100 rpm and the zinc deposits also did not overflow at 150 rpm. Pounder could also prevent the dendrite growth to some extent but it had some difficulties in operation compared with other types of stirrers.