Rapid separation of zirconium using microvolume anion-exchange cartridge for Zr determination with isotope dilution ICP-MS

Asai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*

Talanta, 185, p.98 - 105, 2018/08

https://doi.org/10.1016/j.talanta.2018.03.026

Estimating the risks associated with radiation from long-lived fission products (LLFP) in radioactive waste is essential to ensure the long-term safety of potential disposal sites. In this study, the amount of Zr, a LLFP, was determined by ICP-MS after separating Zr from a spent nuclear fuel solution using a microvolume anion-exchange cartridge (TEDA cartridge). The TEDA cartridge achieved highly selective separation of Zr regardless of its small bed volume of 0.08 cm. The time taken to complete the Zr separation was 1.2 min with a flow rate of 1.5 mL/min, which was 10 times faster than that for a conventional anion-exchange resin column. Almost all the other elements were removed, leading to accurate measurement of Zr. The result connects experimental value to theoretical prediction provided by ORIGEN2, which requires verification. With the measured value, we demonstrated that the theoretical value is reliable enough to estimate radiation risks.

Speciation of Eu(III) in an anion exchange separation system with LiCl-HO/alcohol mixed media by time-resolved laser-induced fluorescence spectroscopy

Arisaka, Makoto*; Kimura, Takaumi; Suganuma, Hideo*; Yoshida, Zenko

Journal of Radioanalytical and Nuclear Chemistry, 255(2), p.385 - 389, 2003/02

https://doi.org/10.1023/A:1022529525952

no abstracts in English

Analysis of Cm contained in irradiated fuel of experimental fast reactor "JOYO"; Development of the analytical technique and measurement of Cm

Osaka, Masahiko; Koyama, Shinichi; Mitsugashira, Toshiaki; Morozumi, Katsufumi; Namekawa, Takashi

JNC TN9400 2000-058, 49 Pages, 2000/04

JNC-TN9400-2000-058.pdf:1.22MB

The analytical technique for Cm contained in a MOX FUEL was developed and analysis of Cm contained in irradiated fuel of experimental fast reactor "JOYO" was carried out, to contribute to evaluation of transmutation characteristics of MA nuclide in the fast reactor. The procedure of ion-exchange separation of Cm with nitric acid-methanol mixed media essential for the isotopic analysis in irradiated MOX fuel was adopted considering for being rapid and easy. The fundamental test to grasp separation characteristics of this procedure, such as Cm elution position and separation capacity between Cm and Am or Eu, was carried out. ln applying this procedure to the analysis of Cm contained in actual specimen, separation condition was evaluated and optimized, and the procedure consist of impurity removal and Am removal process was devised. This procedure resulted in high recovery rate of Cm and high removal rate of Am and impurity which becomes a problem in sample handling and mass-spectrometry such as Eu and Cs. The Cm separation test from irradiated MOX fuel was carried out using this technique, and Cm isotopic ratio analysis was enabled. The analytical technique for Cm contained in irradiated MOX fuel was established using the procedure of ion-exchange separation with nitric acid-methanol mixed media. The analysis of Cm contained in irradiated MOX fuel of experimental fast reactor "Joyo" was carried out. As a result, it was revealed from measured data that Cm content rate was 1.4 4.0lO atom%, small amount of Cm was generated and Cm isotopic ratio was constant above burn-up 60GWd/t.

None

*

PNC TJ6614 93-001, 32 Pages, 1993/03

PNC-TJ6614-93-001.pdf:0.61MB

None

Separation of technetium with active carbon

; Kubota, Masumitsu

Journal of Nuclear Science and Technology, 26(11), p.1038 - 1044, 1989/11

https://doi.org/10.3327/jnst.26.1038

no abstracts in English

Anion exchange separation of hard acid-metals using HF-HBO media

Bulletin of the Chemical Society of Japan, 55, p.1824 - 1830, 1982/00

https://doi.org/10.1246/bcsj.55.1824

no abstracts in English

Amphoteric character of cadmium hydroxide and its solubility in alkaline solution

;

Journal of Inorganic and Nuclear Chemistry, 35(7), p.2592 - 2594, 1973/07

https://doi.org/10.1016/0022-1902(73)80336-7

no abstracts in English

The separation of strontium-90 and yttrium-90 by the oxalic acid type anion exchange resin

; ; *

Bunseki Kagaku, 7, P. 720, 1958/00

no abstracts in English

Study on Sm/Pm separation for spent fuel composition analysis

Sato, Makoto; Ichimura, Seiji; Suyama, Kenya; Tonoike, Kotaro; Kamohara, Keiko*; Suzuki, Norihisa*

no journal, , 

no abstracts in English

Determination of Zr isotopes in spent nuclear fuel with ICP-MS

Asai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi

no journal, , 

Various Zr isotopes generated by U fission are found in spent fuel and HLW. Among them, Zr which has a long half-life of 1.510y has a potential to contribute to radiation dose over an extended period of time after the implementation of HLW disposal. The inventory estimation of Zr in HLW confirmed by measured data is the key to realize a safe and cost-efficient disposal. In this study, a simple and robust analytical technique for the determination of Zr based on ion-exchange chromatography combined with ICP-MS was developed. Interference-free measurement was achieved by a single anion-exchange step, removing Sr, Nb, and Mo which would cause spectral interferences. Additionally, major component U and radioactive components, such as Cs, Ba, and Pu, were also removed concurrently. Concentration of Zr was readily calculated with measured isotope ratios of Zr/Zr in the sample and natural Zr-spiked sample with sufficient accuracy.