• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.245-256
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• 2004

The extraction of three kinds of amido podands, N,N,N'N'-tetrabutyl-3-oxa-pentanedi- amide (TBDGA), N,N,N'N'-tetra-isobutyl-3-oxa-pentanediamide(TiBDGA) and N,N,N'N'-tetra- butyl-3,6-dioxa-oct-anediam- ide(TBDOODA) on U(VI),Pu(IV), Am(III), Eu(III) and other metal ions is studied in nitric acid solutions. 40%octanol-kerosene is chosen as diluents to eliminate third phase and emulsion. TBDGA and TiBDGA show extraction selectivity to An(III) and Ln(III) much higher than to U(VI) and Pu(IV). Fe, Ru and Mo is poorly extracted by the three kinds of amid podands in 2~3mol/L $HNO_3$ solutions. Aiming to eliminate interface crude when using simulated HLLW solution in the system of 0.2mol/L TBDGA/Octanol＋kerosene, acetohydroxyamic acid was adapted. Distribution ratio of zirconium was decreased when adding acetohydroxyamic acid in aqueous solution, and interface crude disappeared as mixing extractant with HLLW. The counter-current extraction test is carried out in a set of miniature mixer-settler, with 0.2mol/L TBDGA/ 40% octanol-kerosene as extractant to separate U(VI), Pu(IV), Am(III) and Eu(III) from simulated high level liquid waste(HLLW) solution. In battery A, lanthanides and actinides are coextracted into organic phase with the recovery of 99.98% for U(Ⅵ), >99.99% for Pu(IV), and >99.99% for Am(III) and Eu(III) respectively. In battery R1, 99.99% U, 86.2% Pu and a part of Am or Eu are stripped into aqueous phase by 0.2mol/L acetohydroxyamic acid (AHA) in 0.01mol/L $HNO_3$ solution. In battery $R_2$, Am, Eu and remained Pu are completely back-extracted by 0.2mol/L AHA. This separation process contains no salt reagent, and it is not necessary to dilute HLLW feed.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.149-158
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• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1799-1804
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• 2019

The concept of a new method, the CARBEX (CARBonate EXtraction) process, was proposed for reprocessing of spent uranium oxide fuel. The proposed process is based on use of water solutions of $Na_2CO_3$ or $(NH_4)_2CO_3$ and solvent extraction (SE) by the quaternary ammonium compounds for selective recovery and purification of U from the fission products (FPs). Applying of SE allows to reach high degree of purification of U from FPs. Carrying out the processes in poorly aggressive alkaline carbonate media leads to increasing safety of SNF's reprocessing and better selectivity of separation of lanthanides and actinides. Moreover carbonate reprocessing media allows to carry out a recycling and regeneration of reagents. We have been done laboratory scale experiments on the extraction components of simulated voloxidated spent fuel in the solutions of NaOH or $Na_2CO_3-H_2O_2$ and recovery of U from carbonate solutions by SE method using carbonate of methyltrioctylammonium in toluene. It was shown that the purification factors of U from impurities of simulated FPs reached values $10^3-10^5$. The received results support our opinion that CARBEX after the further development can become more safe, simple and profitable method of spent fuel reprocessing.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.95-102
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• 2006

Alpha spectrometry was studied for the determination of $^{237}Np$ in spent nuclear fuel samples. The optimum condition for the electrodeposition of $^{237}Np$ was obtained as follows : for $1{\sim}1.5$ hour of deposition time, at the current intensity of $1.2{\sim}1.5$ A and at sodium sulfate electrolyte without organic additive. The deposition yield and its reproducibility on $^{237}Np$ was decreased as the amount of $^{237}Np$ decreased from 4.16 Bq down to 0.0264 Bq(1ng). The recovery yield of $^{237}Np$ determined by alpha spectrometry after separation in synthetic solution was $98.8{\pm}5.1%$(n=4). The contents of $^{237}Np$ in spent nuclear fuel samples were determined and the result showed an agreement within 10% of a difference between the measurement and the calculation.

• Applied Chemistry for Engineering
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• v.7 no.1
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• pp.153-161
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• 1996

Several main nuclides($^{241}Am$, $^{152}Eu$ and $^{237}Np$) in radioactive waste solution were selected and examined to mutual separation with di-(2-ethylhexyl) phosphoric acid by solvent extraction technique. $^{237}Np$ was extracted more than 99.9% adding the $H_2O_2$ that was a good reductant for the oxidation state control of $^{237}Np$. $^{241}Am$, $^{152}Eu$ and $^{237}Np$ could be fairly well separated one another during the different sequence stripping stages, but about 7~9.6% of the other nuclides were still remained for the $^{241}Am$ stripping solution. This result shows that the product of $^{152}Eu$ and $^{237}Np$ was good, but $^{241}Am$ may be needed to further purification process. It was also discussed on the cause of the third phase formation phenomenon that was found in the solvent regeneration.