• Nuclear Engineering and Technology
• /
• v.54 no.9
• /
• pp.3235-3241
• /
• 2022

The Korea Atomic Energy Research Institute (KAERI) is planning the construction of the KIJANG Research Reactor (KJRR) for stable supply of ⁹⁹Mo. The Fission ⁹⁹Mo Production Process (FMPP) of KJRR produces solid waste such as spent uranium cake and alumina cake, and liquid waste in the form of intermediate level liquid waste (ILLW) and low level liquid waste (LLLW). This study thus established the operating range and optimum operating conditions for the cementation of ILLW from FMPP. It also evaluated whether cement waste form samples produced under optimum operational conditions satisfy the waste acceptance criteria (WAC) of a disposal facility in Korea (Korea radioactive waste agency, KORAD). Considering economic feasibility and safety, optimum operational conditions were achieved at a w/c ratio of 0.55, and the corresponding salt content was 5.71 wt%. The cement waste form samples prepared under optimum operational conditions were found to satisfy KORAD's WAC when tested for structural stability and leachability. The results indicate that the proposed cementation conditions for the disposal of ILLW from FMMP can be effectively applied to KJRR's disposal facility.

• 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).

• Journal of Environmental Science International
• /
• v.17 no.10
• /
• pp.1129-1137
• /
• 2008

We investigated the ecological characteristics of reed populations growing in Korea and tried to select reed populations showing better growth patterns in waste landfill leachate. To examine the growth characteristics, 14 reed populations from various habitats were collected. Four reed populations were from inland reclaimed habitats, 4 reed populations from brackish or salt marsh habitats, and 6 reed populations from fresh water habitats. Total plant biomass after the treatment with landfill leachate showed that Daebudo and Nanjido reed populations had the higher biomass with 3755 g DW/pot and 3305 g DW/pot, respectively. Reed populations being sampled from the higher salinity and landfill habitats had relatively higher total biomass than that of other reed populations. Especially reed populations from landfill habitats showed higher biomass. Reed populations from Songjiho and Daebudo, which were believed to have tolerance to salt stress, also showed good growth patterns. Population from the fresh water habitats exhibited relatively lower tolerance to leachate treatment compared to others. From the results, we could conclude that reed populations from Nanjido and Daebudo with higher biomass and better salt tolerance were able to good candidates for purification of waste landfill leachate.

• Korean Journal of Materials Research
• /
• v.26 no.12
• /
• pp.681-687
• /
• 2016

Zirconium(Zr) nuclear fuel cladding tubes are made using a three-time pilgering and annealing process. In order to remove the oxidized layer and impurities on the surface of the tube, a pickling process is required. Zr is dissolved in HF and $HNO_3$ mixed acid during the process and pickling waste acid, including dissolved Zr, is totally discarded after being neutralized. In this study, the waste acid was recycled by adding $BaF_2$, which reacted with the Zr ion involved in the waste acid; $Ba_2ZrF_8$ was subsequently precipitated due to its low solubility in water. It is very difficult to extract zirconium from the as-recovered $Ba_2ZrF_8$ because its melting temperature is $1031^{\circ}C$. Hence, we tried to recover Zr using an electrowinning process with a low temperature molten salt compound that was fabricated by adding $ZrF_4$ to $Ba_2ZrF_8$ to decrease the melting point. Change of the Zr redox potential was observed using cyclic voltammetry; the voltage change of the cell was observed by polarization and chronopotentiometry. The structure of the electrodeposited Zr was analyzed and the electrodeposition characteristics were also evaluated.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.4
• /
• pp.303-309
• /
• 2010

For the pyrochemical process of spent nuclear fuels, recovery of LiCl-KCl eutectic salts is needed to reduce radioactive waste volume and to recycle resource materials. This paper is about recovery of residual LiCl-KCl eutectic salts in radioactive rare earth precipitates (rare earth oxychlorides or oxides) by using a vacuum distillation process. In the vacuum distillation test apparatus, the salts in the rare earth precipitates were vaporized and were separated effectively. The separated salts were deposited in three positions of the vacuum distillation test apparatus or were collected in the filter and it is difficult to recover them. To resolve the problem, a vacuum distillation and condensation system, which is subjected to the force of a temperature gradient at a reduced pressure, was developed. In a preliminary test of the vacuum distillation/condensation recovery system, it was confirmed that it was possible to condense the vaporized salts only in the salt collector and to recover the condensed salts from the salt collector easily.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.20 no.4
• /
• pp.469-488
• /
• 2022

Technology for high-level-waste disposal employing a multibarrier concept using engineered and natural barrier in stable bedrock at 300-1,000 m depth is being commercialized as a safe, long-term isolation method for high-level waste, including spent nuclear fuel. Managing heat generated from waste is important for improving disposal efficiency; thus, research on efficient heat management is required. In this study, thermal management methods to maximize disposal efficiency in terms of the disposal area required were developed. They efficiently use the land in an environment, such as Korea, where the land area is small and the amount of waste is large. The thermal effects of engineered barriers and natural barriers in a high-level waste disposal repository were analyzed. The research status of thermal management for the main bedrocks of the repository, such as crystalline, clay, salt, and other rocks, were reviewed. Based on a characteristics analysis of various heat management approaches, the spent nuclear fuel cooling time, buffer bentonite thermal conductivity, and disposal container size were chosen as efficient heat management methods applicable in Korea. For each method, thermal analyses of the disposal repository were performed. Based on the results, the disposal efficiency was evaluated preliminarily. Necessary future research is suggested.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.06a
• /
• pp.339-339
• /
• 2004

원자로를 이용하여 장수명핵종(long lived nucleus)을 소멸처리하는 과정에서 초우라늄(TRU, transuranium)과 희토류(RE, rare earth) 금속에 포함되어 있는 소량의 핵분열성(fissile) 물질인 우라늄을 제거할 필요가 있다. 본 실험은 LiCl-KCl 용융염계에서 전해제련법(Electrowinning)을 이용하여 용융염욕에 존재하는 우라늄을 제거하기 위하여 필요한 Cd-Li 양전극 물질을 제조하였고, 제조된 금속을 이용하여 우라늄 및 란탄족(Dy, Ce, Y, Nd, Gd) 금속의 환원 특성을 파악하였다.(중략)

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.11a
• /
• pp.145-147
• /
• 2004

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.11a
• /
• pp.581-581
• /
• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2017.05a
• /
• pp.227-228
• /
• 2017

In this paper, electrolytic behavior of Cerium and Ce-Bi ion system was studied. The electrochemical behavior of Ce was studied in $LiCl-KCl-CeCl_3$ molten salts using electrochemical techniques Cyclic Voltammetry on tungsten electrodes at 773K. During the process of CV electrolysis, intermetallic compound were observed of Ce, Cex-Biy. Further study, in order to determine clarity of diffusion coefficient in this experiment, we will compare result of electrochemistry method and we also need to quantitative research.