Study on gamma-ray-degradation of adsorbent for low pressure-loss extraction chromatography

Miyazaki, Yasunori; Sano, Yuichi; Okamura, Nobuo; Watanabe, Masayuki; Koka, Masashi*

QST-M-29; QST Takasaki Annual Report 2019, P. 72, 2021/03

https://repo.qst.go.jp/records/82107

no abstracts in English

Recent trend of the radionuclide analyses in bioassay

Tomita, Jumpei

Bunseki, 2019(3), p.112 - 113, 2019/03

no abstracts in English

Effect of surface treatment on adsorption and desorption behavior of impregnated adsorbent for extraction chromatography

Nagoshi, Kohei*; Arai, Tsuyoshi*; Watanabe, So; Sano, Yuichi; Takeuchi, Masayuki; Sato, Mutsumi*; Oikawa, Hiroshi*

Nihon Ion Kokan Gakkai-Shi, 28(1), p.11 - 18, 2017/01

https://doi.org/10.5182/jaie.28.11

no abstracts in English

Synthesis and detection of a Seaborgium carbonyl complex

Even, J.*; Yakushev, A.*; Dllmann, Ch. E.*; Haba, Hiromitsu*; Asai, Masato; Sato, Tetsuya; Brand, H.*; Di Nitto, A.*; Eichler, R.*; Fan, F. L.*; et al.

Science, 345(6203), p.1491 - 1493, 2014/09

https://doi.org/10.1126/science.1255720

A new superheavy element complex, a seaborgium carbonyl, has been successfully synthesized, and its adsorption property has been studied using a cryo-thermochromatography and -detection apparatus COMPACT. Nuclear reaction products of short-lived Sg preseparated with a gas-filled recoil ion separator GARIS at RIKEN were directly injected into a gas cell filled with He/CO mixture gas, and chemical reaction products of volatile carbonyl complexes were trasported to COMPACT. The Sg carbonyl complex detected with COMPACT was found to be very volatile with adsorption enthalpy of 50 kJ/mol, from which we have concluded that this complex should be a Sg hexacarbonyl Sg(CO). This is the first synthesis of organometallic compounds of transactinide elements for which only simple inorganic comounds have been synthesized so far.

Separation of trivalent actinides from lanthanides by using R-BTP resins and stability of R-BTP resin

Hoshi, Harutaka*; Wei, Y.*; Kumagai, Mikio*; Asakura, Toshihide; Morita, Yasuji

Journal of Alloys and Compounds, 408-412, p.1274 - 1277, 2006/02

https://doi.org/10.1016/j.jallcom.2005.04.128

For the development of advanced aqueous reprocessing system, it is one of the most important subjects to separate minor trivalent actinides (MA = Am and Cm). Recently, extraction selectivity for MA(III) over Ln(III) has been found in some extractants containing soft donor, such as S or N, ligands. Kolarik et al. reported that a new N-donor ligand 2,6-bis(5,6-dialkyl-1,2,4-triazine-3-yl)-pyridine (R-BTP) shows high selectivity for MA (III) over Ln(III). The novel silica-based extraction resins were prepared by impregnating some R-BTP molecules into a macroreticular styrene-divinylbenzene copolymer which is immobilized in porous silica particles with a mean diameter of 50 m. Separation of simulated high level liquid waste solution containing Ln(III) and trace amount of Am(III) was studied. Am(III) was mutually separated from Ln(III) through a packed column with R-BTP impregnating resin, very high decontamination factor ( 10) for Am, and all the elements were recovered quantitatively.

Analysis of Re-EDTMP complexes by HPLC and ultrafiltration

Hashimoto, Kazuyuki; Matsuoka, Hiromitsu

Radiochimica Acta, 92(4-6), p.285 - 290, 2004/07

https://doi.org/10.1524/ract.92.4.285.35587

no abstracts in English

Production of no-carrier-added Lu via the Yb(n,)YbLu process

Hashimoto, Kazuyuki; Matsuoka, Hiromitsu; Uchida, Shoji*

Journal of Radioanalytical and Nuclear Chemistry, 255(3), p.575 - 579, 2003/03

https://doi.org/10.1023/A:1022557121351

The emitter Lu is a promising therapeutic radioisotope for the treatment of cancer. It has a half-life of 6.73 days and maximum energy of 498 keV, resulting in a short range of radiation in tissue. The decay is accompanied by the emission of low energy -radiation with = 208 keV (11.0%) and 113 keV (6.4%) suitable for simultaneous imaging. Lutetium-177 can be usually produced at nuclear reactors with high yield and high specific radioactivity by the Lu(n,)Lu reaction. However, radioisotopes with higher specific radioactivity are required in the field of radioimmunotherapy using labeled monoclonal antibodies. Thus, an alternative production route, namely the Yb(n,)Yb Lu process was studied to produce no-carrier-added (nca) Lu in this work. The radiochemical separation of the nca Lu from the macroscopic ytterbium target was investigated by means of reversed-phase ion-pair HPLC. The nca Lu was obtained in radiochemical pure form with a separation yield of 80%.

A Study on modeling and numerical simulation of extraction in the CMPO-TBP system

; ;

JNC TN8400 2001-022, 60 Pages, 2001/03

JNC-TN8400-2001-022.pdf:1.31MB

A numerical simulation code for the TRUEX (Transuranium Extraction) process was developed. Concentration profiles of americium and europium were calculated for some experiments of the counter current extraction system those were carried out in CPF (Chemical Processing Facility) by using the code. Calculation profiles were in agreement with the experimental results. Operational conditions were also examinted for the americium recovery experiment by the TRUEX process carried out in the Plutonium Fuel Center. It was shown that lowering the concentration of nitric acid in the scrub solution and decreasing the flow rate of solvent and strip solution was effective for improving the performance of the stripping step and reducing the volume of the waste solution. In order to find the optimum conditions for various experiments, this simulation code was modified to calculate the concentration profiles of other metal elements such as zirconium and iron and the effect of oxalic acid on the extraction behavior of the metal elements. The calculated concentration profiles of americium and europium were varied by this modification. In the experiment at CPF, the calculations were carried out to obtain recovery ratio of americium in the product stream with the amount of oxalic acid added to the process. This calculation result showed that it was possible to improve the performance of decontamination of fission products by increasing oxalic acid concentration added to the process. The calculation was also carried out for finding the optimum conditions of oxalic acid concentration added to the europium recovery process.

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.

A Study on the reprocessing of spent FBR-fuel by ion exchange

*; Arai, Tsuyoshi*; Kumagai, Mikio*

JNC TJ9400 2000-002, 80 Pages, 2000/02

JNC-TJ9400-2000-002.pdf:4.67MB

In order to develop an economically efficient wet separation process other than solvent extraction for reprocessing spent FBR-fuel (MOX fuel), we have investigated the possibility of an advanced ion exchange process. Based on the fundamental research results, we proposed an advanced ion exchange process considering the characteristics of FBR-fuel cycle. The separation system consists of a main separation process using a novel anion exchanger which has a rapid kinetics and two extraction chromatography processes for minor actinides isolation using novel impregnation adsorbents with high selectivity. The chemical flow sheet, mass balance chart, list of main equipment and installation layout of each equipment were estimated and designed for the process in a reprocessing plant with the capacity of 200 tHM/y FBR-fuel. The process was pfeliminarny evalualed from the aspects of economy performance, recovery of potentially useable resources, minimization of environmental risk and proliferation-resistance by comparing with the advanced PUREX process. Furthermore, the subjects which are important for the practical application of the process are also listed.