• Journal of the Korean institute of surface engineering
• /
• v.51 no.1
• /
• pp.34-39
• /
• 2018

Steel crucible used for molten Al has a problem of very limited lifetime because of the interaction between Fe and molten Al. This study was performed to improve the lifetime of steel crucible for molten Al by coating metallic Al and by further anodizing treatment to form thick and uniform anodic oxide films. The lifetime of the steel crucible was improved slightly by Al coating from 30 to 40 hours by metallic Al coating and largely to 120 hours by coating the surface with anodic oxide film. The improved lifetime was attributed to blocking of the reaction between Fe and molten Al with the help of anodic oxide layer with more than 20 um thickness on the crucible surface. The failure of the steel crucible arises from the formation of intermetallic compounds and pores at the steel/Al interface.

• Journal of Korea Foundry Society
• /
• v.35 no.5
• /
• pp.114-119
• /
• 2015

This paper reports the effect of a refractory crucible type on the microstructure of duplex stainless steel (DSS) cast with the addition of gadolinium using air-induction melting. Grade 4A DSSs with 1 wt% of gadolinium (Gd) were fabricated in various crucibles including alumina ($Al_2O_3$), magnesia (MgO), calcia (CaO) coated with yttria ($Y_2O_3$) and graphite. The standard free energies of the formation of calcium and yttrium oxide were lower than those of gadolinium oxide and other crucible elements based oxide. The yield of Gd in DSS using $Al_2O_3$, MgO, CaO-coated $Y_2O_3$ and graphite was 5, 19, 83 and 96%, respectively. As Gd yield increased, the amount of Gd-based inclusions increased, the size of the inclusions were reduced, and the inclusions became evenly distributed.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.05a
• /
• pp.57-57
• /
• 2018

The electrorefining process is generally composed of two recovery steps in pyroprocessing - the deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The liquid cathode processing is necessary to separate cadmium from the actinide elements since the actinide deposits are dissolved or precipitated in a liquid cathode. Distillation process was employed for the cathode processing. It is very important to avoid a splattering of cadmium during evaporation due to the high vapor pressure. In this study, a multi-layer porous round cover was proposed and examined to develop a splatter shield for the Cd distillation crucible. Cadmium vapor can be released through the holes of the shield, whereas liquid drops can be collected in the multiple hemisphere. The collected drops flow on the round surface of the cover and flow down into the crucible. The crucible cover was fabricated and tested in the Cd distiller. The cover was made with three stainless steel round plates with a diameter of 33.50 mm. The distance between the hemispheres and the diameter of the holes are 10 and 1 mm, respectively. About 40 grams of Cd and about 4 grams of Bi was distilled at a reduced pressure for two hours at $470^{\circ}C$. After the Cd distillation experiment, cadmium was not detected and more than 90 % of Bi remained in the ICP-OES analysis. Therefore the crucible cover can be a candidate for the splatter shield of the Cd distillation crucible. Further development of the crucible cover is necessary for the decision of the optimum cover geometry and the operating conditions of the Cd distiller.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.11a
• /
• pp.103-103
• /
• 2018

The liquid cathode processing is necessary to separate cadmium from the actinide elements in the pyroprocessing since the actinide deposits are dissolved or precipitated in a liquid cathode. Distillation process was employed for the cathode processing owing to the compactness. It is very important to avoid a splattering of cadmium during evaporation due to the high vapor pressure. Several methods have been proposed to lower the splattering of cadmium during distillation. A multi-layer porous round cover was proposed to avoid a cadmium splattering in our previous study. In this study, distillation behavior of $Cd-ZrO_2$ and Cd - Bi systems were investigated to examine a multi-layer porous round cover for the development of the cadmium splatter shield of distillation crucible. It was designed that the cadmium vapor can be released through the holes of the shield, whereas liquid drops can be collected in the multiple hemisphere. The cover was made with three stainless steel round plates with a diameter of 33.50 mm. The distance between the hemispheres and the diameter of the holes are 10 and 1 mm, respectively. Bismuth or zirconium oxide powder was used as a surrogate for the actinide elements. About 40 grams of Cd was distilled at a reduced pressure for two hours at various temperatures. The mixture of the cadmium and the surrogate was distilled at 470, 570 and $620^{\circ}C$ in the crucible with the cover. Most of the bismuth or zirconia remained in the crucible after distillation at 470 and $570^{\circ}C$ for two hours. It was considered that the crucible cover hindered the splattering of the liquid cadmium from the distillation crucible. A considerable amount of the surrogate material reduced after distillation at $620^{\circ}C$ due to the splattering of the liquid cadmium. The low temperature is favorable to avoid a liquid cadmium splattering during distillation. However, the optimum temperature for the cadmium distillation should be decided further, since the evaporation rate decreases with a decreasing temperature.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2017.10a
• /
• pp.73-73
• /
• 2017

Pyroprocessing at KAERI (Korea Atomic Energy Research Institute) consists of pretreatment, electroreduction, electrorefining and electrowinning. SFR (Sodium Fast Reactor) fuel is prepared from the electrowinning process which is composed of LCC (Liquid Cadmium Process) and Cd distillation et al. LCC is an electrochemical process to obtain actinides from spent fuel. In order to recover actinides inert anodes such as carbon material are used, where chlorine gas ($Cl_2$) evolves on the surface of the carbon material. And, stainless steel (SUS) crucible should be installed in large-scale electrowinning system. Therefore, the effect of chlorine on the SUS material needs to be studied. LiCl-KCl-$UCl_3$-$NdCl_3$-$CeCl_3$-$LaCl_3$-$YCl_3$ salt was contained in 2 kinds of electrolytic crucible having an inner diameter of 5cm, made of an insulated alumina and an SUS, respectively. And, three kinds of electrodes such as cathode, anode, reference were used for the electrochemical experiments. Both solid tungsten (W) and LCC were used as cathodes. Cd of 45 g as the cathode material was contained in alumina crucibles for the deposition experiments, where the crucible has an inner diameter of 3 cm. Glassy carbon rod with the diameter of 0.3 cm was employed as an anode, where shroud was not used for the anode. A pyrex tube containing LiCl-KCl-1mol% AgCl and silver (Ag) wire having a diameter of 0.1cm was used as a reference electrode. Electrodeposition experiments were conducted at $500^{\circ}C$ at the current densities of $50{\sim}100mA/cm^2$. In conclusion, Fe ions were produced in the salt during the electrodeposition by the reaction of chlorine evolved from the anode and Fe of the SUS crucible and thereby LCC system using SUS crucible showed very low current efficiencies compared with the system using the insulated alumina crucible. Anode shroud needs to be installed around the glassy carbon not to influence surrounding SUS material.

• Nuclear Engineering and Technology
• /
• v.49 no.7
• /
• pp.1547-1554
• /
• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.

• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1096-1108
• /
• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• Journal of the Korean institute of surface engineering
• /
• v.31 no.6
• /
• pp.371-378
• /
• 1998

The zind bate contaminated in the hot dip galvannealed operation was successfully by appling the dross formation mechanism ; the Fe content was lowered from 0.028% to 0.011% and the dress size was decreased from 15~20$\mu\textrm{m}$ to under 3$\mu\textrm{m}$. The cooled metal from CGL zinc bath during operation of the galvannealed steel strip was remelted in graphite crucible at the lab and agitated after increasing Al content from 0.14% to 0.16% with decreasing the molten metal temperature from $470^{\circ}C$to $445^{\circ}C$. The agitating was done by agitator and nitrogen. The molten was analyed by SEM and EDS. It was considered that the Fe and the bottom dross($FeZN_7$) could react with aluminium to from the float dress($Fe_2Al_5$) according to the molten metal temperature down and the float dress rise to the surface of the zine bath. So the Fe and dross in the bath could be romoved out of the bath. It was confirmed that the proper purication conditions of GA zine bath is 0.02% of Al increasing, bath temperature down from $460^{\circ}C$ to $450^{\circ}C$and agitator and nitrogen.

• Nuclear Engineering and Technology
• /
• v.42 no.2
• /
• pp.131-144
• /
• 2010

The Korea Atomic Energy Research Institute (KAERI) has been developing pyroprocessing technology for recycling useful resources from spent fuel since 1997. The process includes pretreatment, electroreduction, electrorefining, electrowinning, and a waste salt treatment system. This paper briefly addresses unit processes and related innovative technologies. As for the electroreduction step, a stainless steel mesh basket was applied for adaption of granules of uranium oxide. This basket was designed for ready handling and transfer of feed material. A graphite cathode was used for the continuous collection of uranium dendrite in the electrorefining system. This enhances the throughput of the electrorefiner. A particular mesh type stirrer was designed to inhibit uranium spill-over at the liquid Cd crucible. A residual actinide recovery system was also tested to recover TRU tracer. In order to reduce the waste volume, a crystallization method is employed for Cs and Sr removal. Experiments on the unit processes were tested successfully, and based on the results, engineering-scale equipment has been designed for the PRIDE (PyRoprocess Integrated inactive DEmonstration facility).

• Progress in Superconductivity
• /
• v.8 no.1
• /
• pp.98-103
• /
• 2006

[$Bi_2Sr_2Ca_1Cu_2O_x$](BSCCO 2212) tubes were fabricated by centrifugal melting process(CMP). BSCCO 2212 powder was melted in a Pt crucible and solidified in a rotating steel mold. The BSCCO 2212 tube samples separated from steel mold were heat-treated at $800^{\circ}C\;and\;810^{\circ}C$ in oxygen flow. The critical current($I_c$) of the samples was measured by transport measurement at 77K, and the microstructure was investigated by scanning electron microscope. The $J_c$ at 77K of the tubes heated treated at $800^{\circ}C\;and\;810^{\circ}C$ were 197 and 240 $A/cm^2$, respectively. It was observed that the plate like grains in BSCCO 2212 tube was more developed in the sample heat treated at $810^{\circ}C\;and\;800^{\circ}C$ heated tube. It is found that the critical current($I_c$) of the BSCCO 2212 sample was dependent on the heating schedule regarding the growth of the BSCCO 2212 grains.