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      모의 사용후핵연료에 함유된 루테늄의 양이온교환 분리 및 정량 = Cation Exchange Separation and Determination of Ruthenium in a Simulated Spent Nuclear Fuel

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      https://www.riss.kr/link?id=A101032360

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      Cation exchange separation and inductively coupled plasma atomic emission spectrometric(ICP-AES) determination of ruthenium in HCl solutions were studied to quantitatively determine ruthenium in spent nuclear fuels. Ruthenium-bearing samples were dissolved with the mixed acid solution(9 : 1 mole ratio, HCl-HNO$_3$) using an acid digestion bomb. Based on the absorption spectra and ion exchange behaviour of ruthenium in hydrochloric acid media, its possible chemical species were discussed. On a cation exchange column (0.7 ${\times}$ 8.0 cm) packed with AG 50W ${\times}$ 8(100~200 mesh) and equilibrated with 0.5 M HCl, ruthenium was eluated with 0.5 M HCl while uranium was retained on the column. The established separation method was applied to a simulated spent nuclear fuel and resulted in the recovery of 98.5% with a relative standard deviation of 0.7%.
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      Cation exchange separation and inductively coupled plasma atomic emission spectrometric(ICP-AES) determination of ruthenium in HCl solutions were studied to quantitatively determine ruthenium in spent nuclear fuels. Ruthenium-bearing samples were diss...

      Cation exchange separation and inductively coupled plasma atomic emission spectrometric(ICP-AES) determination of ruthenium in HCl solutions were studied to quantitatively determine ruthenium in spent nuclear fuels. Ruthenium-bearing samples were dissolved with the mixed acid solution(9 : 1 mole ratio, HCl-HNO$_3$) using an acid digestion bomb. Based on the absorption spectra and ion exchange behaviour of ruthenium in hydrochloric acid media, its possible chemical species were discussed. On a cation exchange column (0.7 ${\times}$ 8.0 cm) packed with AG 50W ${\times}$ 8(100~200 mesh) and equilibrated with 0.5 M HCl, ruthenium was eluated with 0.5 M HCl while uranium was retained on the column. The established separation method was applied to a simulated spent nuclear fuel and resulted in the recovery of 98.5% with a relative standard deviation of 0.7%.

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