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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
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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
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Computational Analysis for a Molten-salt Electrowinner with Liquid Cadmium Cathode
Kim, Kwang-Rag ; Jung, Young-Joo ; Paek, Seung-Woo ; Kim, Ji-Yong ; Kwon, Sang-Woon ; Yoon, Dal-Seong ; Kim, Si-Hyung ; Shim, Jun-Bo ; Kim, Jung-Gug ; Ahn, Do-Hee ; Lee, Han-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 1~7
In the present work, an electrowinning process in the LiCl-KCl/Cd system is considered to model and analyze the electrotransport of the actinide and rare-earth elements. A simple dynamic modeling of this process was performed by taking into account the material balances and diffusion-controlled electrochemical reactions in a diffusion boundary layer at an electrode interface between the molten salt electrolyte and liquid cadmium cathode. The proposed modeling approach was based on the half-cell reduction reactions of metal chloride occurring on the cathode. This model demonstrated a capability for the prediction of the concentration behaviors, a faradic current of each element and an electrochemical potential as function of the time up to the corresponding electrotransport satisfying a given applied current based on a galvanostatic electrolysis. The results of selected case studies including five elements (U, Pu, Am, La, Nd) system are shown, and a preliminary simulation is carried out to show how the model can be used to understand the electrochemical characteristics and provide better information for developing an advanced electrowinner.
Development of Liquid Cadmium Cathode Structure for the Inhibition of Uranium Dendrite Growth
Paek, Seung-Woo ; Yoon, Dal-Seong ; Kim, Si-Hyung ; Shim, Jun-Bo ; Ahn, Do-Hee ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 9~17
The LCC (Liquid Cadmium Cathode) structure to be developed for inhibiting the formation and growth of the uranium dendrite has been known as a key part in the electrowinning process for the simultaneous recovering of uranium and TRU (TRans Uranium) elements from spent fuels. A zinc-gallium (Zn-Ga) experimental system which is able to be functional in aqueous condition and normal temperature has been set up to observe the formation and growth phenomena of the metal dendrites on liquid cathode. The growth of the zinc dendrites on the gallium cathode and the performance of the existing stirrer type and pounder type cathode structure were observed. Although the mechanical strength of the dendrites appeared to be weak in the electrolyte and easily crashed by the various cathode structures, it was difficult to effectively submerge the dendrite into the bottom of the liquid cathode. Based on the results of the aqueous phase experiments, a lab-scale electrowinning experimental apparatus which are applicable to the development of LCC srtucture for the electrowinning process was established and the performance tests of the different types of LCC structure were conducted to prohibit the uranium dendrite growth on LCC surface. The experimental results of the stirrer type LCC structures have shown that they could not effectively remove the uranium dendrites growing at the inner side of the LCC crucible and the performances of the paddle and harrow type LCC structure were similar. Therefore a mesh type LCC structure was developed to push down the uranium dendrites to the bottom of the LCC crucible growing on the LCC surface and at the inner side of the crucible. From the experimental results for the performance test of the mesh type LCC structure, the uranium was recovered over 5 wt% in cadmium without the growth of uranium dendrites. After completion of the experiments, solid precipitates of the bottom of the LCC crucible were identified as an intermetallic compound (UCd11) by the chemical analysis.
A Chemical Reaction Calculation and a Semi-Empirical Model for the Dynamic Simulation of an Electrolytic Reduction of Spent Oxide Fuels
Park, Byung-Heung ; Hur, Jin-Mok ; Lee, Han-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 19~32
Electrolytic reduction technology is essential for the purpose of adopting pyroprocessing into spent oxide fuel as an alternative option in a back-end fuel cycle. Spent fuel consists of various metal oxides, and each metal oxide releases an oxygen element depending on its chemical characteristic during the electrolytic reduction process. In the present work, an electrolytic reduction behavior was estimated for voloxidized spent fuel based on the assumption that each metal-oxygen system is independent and behaves as an ideal solid solution. The electrolytic reduction was considered as a combination of a Li recovery and chemical reactions between the metal oxides such as uranium oxide and the produced Li metal. The calculated result revealed that most of the metal oxides were reduced by the process. It was evaluated that a reduced fraction of lanthanide oxides increased with a decreasing
concentration. However, most of the lanthanides were expected to be stable in their oxide forms. In addition, a semi-empirical model for describing
electrolytic reduction behavior was proposed by considering Li diffusion and a chemical reaction between
and Li. Experimental data was used to determine model parameters and, then, the model was applied to calculate the reduction yield with time and to estimate the required time for a 99.9% reduction.
Corrosion Behavior of
Coating in an Electrolytic Reduction Process
Cho, Soo-Haeng ; Hong, Sun-Seok ; Kang, Dae-Seung ; Jeong, Myeong-Soo ; Park, Byung-Heong ; Hur, Jin-Mok ; Lee, Han-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 33~39
The electrolytic reduction of a spent oxide fuel involves a liberation of the oxygen in a molten LiCl electrolyte, which results in a chemically aggressive environment that is too corrosive for typical structural materials. Accordingly, it is essential to choose the optimum material for the processing equipment that handles the high molten salt. In this study, hot corrosion studies were performed on bare as well as coated superalloy specimens after exposure to lithium molten salt at
for 216 h under an oxidizing atmosphere. The IN713LC superalloy specimens were sprayed with an aluminized NiCrAlY bond coat and then with an
top coat. The bare superalloy reveals an obvious weight loss due to spalling of the scale by the rapid scale growth and thermal stress. The chemical and thermal stability of the top coat has been found to be beneficial for increasing to the corrosion resistance of the structural materials for handling high temperature lithium molten salts.
Study of the Formation of Eutectic Melt of Uranium and Thermal Analysis for the Salt Distillation of Uranium Deposits
Park, Sung-Bin ; Cho, Dong-Wook ; Hwang, Sung-Chan ; Kang, Young-Ho ; Park, Ki-Min ; Jun, Wan-Gi ; Kim, Jeong-Guk ; Lee, Han-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 41~48
Uranium deposits from an electrorefining process contain about 30% salt. In order to recover pure uranium and transform it into an ingot, the salts have to be removed from the uranium deposits. Major process variables for the salt distillation process of the uranium deposits are hold temperature and vacuum pressure. Effects of the variables on the salt removal efficiency were studied in the previous study. By applying the Hertz-Langmuir relation to the salt evaporation of the uranium deposits, the evaporation coefficients were obtained at the various conditions. The operational conditions for achieving above 99% salt removal were deduced. The salt distilled uranium deposits tend to form the eutectic melt with iron, nickel, chromium for structural material of salt evaporator. In this study, we investigated the hold temperature limitation in order to prevent the formation of the eutetic melt between urnaium and other metals. The reactions between the uranium metal and stainless steel were tested at various conditions. And for enhancing the evaporation rate of the salt and the efficient recovery of the distilled salt, the thermal analysis of the salt distiller was conducted by using commercial CFX software. From the thermal analysis, the effect of Ar gas flow on the evaporation of the salt was studied.
Immobilization of Radioactive Rare Earth oxide Waste by Solid Phase Sintering
Ahn, Byung-Gil ; Park, Hwan-Seo ; Kim, Hwan-Young ; Lee, Han-Soo ; Kim, In-Tae ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 49~56
In the pyroprocessing of spent nuclear fuels, LiCl-KCl waste salt containing radioactive rare earth chlorides are generated. The radioactive rare earth oxides are recovered by co-oxidative precipitation of rare earth elements. The powder phase of rare eath oxide waste must be immobilized to produce a monolithic wasteform suitable for storage and ultimate disposal. The immobilization of these waste developed in this study involves a solid state sintering of the waste with host borosilicate glass and zinc titanate based ceramic matrix(ZIT). And the rare-earth monazite which synthesised by reaction of ammonium di-hydrogen phosphate with the rare earth oxides waste, were immobilzed with the borosilicate glass. It is shown that the developed ZIT ceramic wasteform is highly resistant the leaching process, high density and thermal conductivity.
Reuse Technology of LiCl Salt Waste Generated from Electrolytic Reduction Process of Spent Oxide Fuel
Cho, Yung-Zun ; Jung, Jin-Seok ; Lee, Han-Soo ; Kim, In-Tae ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 57~63
Layer crystallization process was tested for the separation(or concentration) of cesium and strontium fission products in a LiCl waste salt generated from an electrolytic reduction process of a spent oxide fuel. In a crystallization process, impurities (CsCl and
) are concentrated in a small fraction of the LiCl salt by the solubility difference between the melt phase and the crystal phase. Based on the phase diagram of LiCl-CsCl-
system, the separation possibility by using crystallization was determined and the molten salt temperature profile during layer crystallization operation was predicted by using mathematical calculation. In the layer crystallization process, the crystal growth rate strongly affects the crystal structure and therefore the separation efficiency. In the conditions of about 20-25 l/min cooling air flow rate and less than 0.2g/min/
crystal flux, the separation efficiency of both CsCl and
showed about 90% by the layer crystallization process, assuming a LiCl salt reuse rate of 90wt%.
Fundamental Study on a Distillation Separation of a LiCl-KCl Eutectic Salt from Rare Earth Precipitates
Yang, Hee-Chul ; Eun, Hee-Chul ; Kim, In-Tae ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 65~70
The distillation rate on LiCl-KCl eutectic salt under different vacuums from 0.5-50 mmHg was first investigated by using both a non-isothermal and a isothermal thermogravimetric (TG) analysis. Based on the non-isothermal TG data, distillation rate equations as a function of the temperature could be derived. Calculated flux by these model flux equations was in agreement with the distillation rate obtained from isothermal TG analysis. A distillation rate of
is obtainable at temperatures less than 1300K and vacuums of 0.5-50 mmHg. About a 99% salt distillation efficiency was obtained after an hour at a temperature above 1150 K under 50 mmHg in a small scale distillation test system. An increase in the vaporizing surface area is relatively effective for removing residual salt in the remaining particles, when compared to that for the vaporizing time. Over 99.95% of total distillation efficiency was obtained for a 1-h distillation operation by increasing the inner surface area from
The Development of U-recovery by Continuous Electrorefining
Kim, Jeong-Guk ; Park, Sung-Bin ; Hwang, Sung-Chan ; Kang, Young-Ho ; Lee, Sung-Jai ; Lee, Han-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 71~76
The electrorefining process, one of main processes which are composed of pyroprocess to recover the useful elements from spent fuel, and the domestic development of electrorefiner have been reviewed. The electrorefiner is composed of an anode basket containing reduced spent fuel such as uranium, transuranic and rare earth elements, and a solid cathode, which are in LiCl-KCl eutectic electrolyte. Oxidation (dissolution) reaction occurs on the anode and a pure uranium is electrochemically reduced (deposited) on the solid cathode. By application of graphite cathode, which has a self-scrapping characteristics for the uranium deposits, and a recovery of the fallen deposits by a screw conveyer, a high-throughput continuous electrorefiner with a capacity of 20 kgU/day has been developed.
Development of an Oxide Reduction Process for the Treatment of PWR Spent Fuel
Hur, Jin-Mok ; Hong, Sun-Seok ; Jeong, Sang-Mun ; Lee, Han-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 77~84
Reduction of oxides has been investigated for the volume reduction and recycling of the spent oxide fuel from commercial nuclear power plants. Various oxide reduction methods were proposed and KAERI (Korea Atomic Energy Research Institute) is currently developing an electrochemical reduction process using a LiCl-
molten salt as a reaction medium. The electrochemical reduction process, the front end of the pyroprocessing, can connect the PWR (Pressurized Water Reactor) oxide fuel cycle to a metal fuel cycle of the sodium cooled fast reactor. This paper summarizes KAERI efforts on the development, improvement, and scale-up of the oxide reduction process.
Uranium ingot casting method with Uranium deposit in a Pyroprocessing
Lee, Yoon-Sang ; Cho, Choon-Ho ; Lee, Sung-Ho ; Kim, Jeong-Guk ; Lee, Han-Soo ;
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT), volume 8, issue 1, 2010, Pages 85~89
The uranium ingot casting process is one of the steps which consolidate uranium deposits produced by electrorefiner as an ingot form in a pryprocessing technique. This paper introduces new design concept of the ingot casting equipment and the performance test results of the lab-scale ingot casting equipment fabricated based on the design concept. Casting equipment produces the uranium ingot by pouring an uranium melt into a mold by tilting a melting crucible. Also it is equipped with a cup which is able to continuously feed uranium deposits into a melting crucible. The productivity could be significantly enhanced by introducing the continuous operation concept.