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Sr in urine sample; Rapid separation of Sr by phosphate co-precipitation and extraction chromatography, followed by determination by triple quadrupole inductively coupled plasma mass spectrometry (ICP-MS/MS)Tomita, Jumpei; Takeuchi, Erina
Applied Radiation and Isotopes, 150, p.103 - 109, 2019/08
Times Cited Count：1 Percentile：53.3(Chemistry, Inorganic & Nuclear)A rapid analytical method for determining Sr in urine samples (1-2 L) was developed to assess the internal exposure of workers in a radiological emergency. Strontium in a urine sample was rapidly separated by phosphate co-precipitation, followed by extraction chromatography with a tandem column of Pre-filter, TRU and Sr resin, and the Sr activity was determined by ICP-MS/MS. Measurement in the MS/MS mode with an O
reaction gas flow rate 1 mL min
showed no tailing of 
Sr at m/z = 90 up to 50 mg-Sr L. The interferences of Ge, Se and Zr at m/z = 90 were successfully removed by chemical separation. This analytical method was validated by the results of the analyses of synthetic urine samples (1.2-1.6 L) containing a known amount of Sr along with 1 mg of each of Ge, Se, Sr and Zr. The turnaround time for analysis was about 10 h, and the detection limit of Sr was approximately 1 Bq per urine sample.
Tomita, Jumpei
Bunseki, 2019(3), p.112 - 113, 2019/03
no abstracts in English
-P adsorbent on extraction chromatography processWatanabe, So; Sano, Yuichi; Sanda, Shuhei*; Sakurai, Shota*; Arai, Tsuyoshi*
Nippon Ion Kokan Gakkai-Shi, 30(1), p.8 - 16, 2019/01
Watanabe, So; Suzuki, Hideya; Goto, Ichiro*; Kofuji, Hirohide; Matsumura, Tatsuro
Nippon Ion Kokan Gakkai-Shi, 29(3), p.71 - 75, 2018/09
Sano, Yuichi; Watanabe, So; Nakahara, Masaumi; Aihara, Haruka; Takeuchi, Masayuki
Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL 2017) (USB Flash Drive), 4 Pages, 2017/09
JAEA has been promoting MA recycle project using a FR fuel cycle named as SmART cycle concept. The SmART cycle contains the recovery of all actinides, in which total amount of MA is estimated to around 1-2g, at CPF from the FR Joyo spent fuel, the fabrication of MA bearing MOX fuel pellets and pins at AGF with recovered actinides, and the irradiation test of the fabricated fuels at the Joyo. In this paper, recent activities on actinides recovery in CPF, which will make a significant contribution to the SmART cycle, were summarized.
Watanabe, So; Sano, Yuichi; Shiwaku, Hideaki; Yaita, Tsuyoshi; Ono, Shimpei*; Arai, Tsuyoshi*; Matsuura, Haruaki*; Koka, Masashi*; Sato, Takahiro*
Nuclear Instruments and Methods in Physics Research B, 404, p.202 - 206, 2017/08
Times Cited Count：3 Percentile：48.33(Instruments & Instrumentation)Nagoshi, Kohei*; Arai, Tsuyoshi*; Watanabe, So; Sano, Yuichi; Takeuchi, Masayuki; Sato, Mutsumi*; Oikawa, Hiroshi*
Nippon Ion Kokan Gakkai-Shi, 28(1), p.11 - 18, 2017/01
no abstracts in English
Hoshi, Harutaka*; Wei, Y.*; Kumagai, Mikio*; Asakura, Toshihide; Morita, Yasuji
Recent Advances in Actinide Science, p.596 - 598, 2006/06
Recently, extraction selectivity for trivalent minor actinides (MA = Am and Cm) over lanthanides (Ln) 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) [1]. However, protonation of R-BTP results in its acidic hydrolysis in acidic medium. Stability in acidic solution was improved by substitution of long normal chain or branched chain [2]. In this work, separation of MA(III) and Ln(III) from nitric acid solution was studied by using novel R-BTP impregnated resin. Branched R-BTP resin had high affinity for Am from up to 4 M HNO
solution and its distribution coefficient was over 10
.
Hoshi, Harutaka*; Wei, Y.*; Kumagai, Mikio*; Asakura, Toshihide; Morita, Yasuji
Journal of Alloys and Compounds, 408-412, p.1274 - 1277, 2006/02
Times Cited Count：36 Percentile：13.61(Chemistry, Physical)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.
Hoshi, Harutaka*; Wei, Y.*; Kumagai, Mikio*; Asakura, Toshihide; Morita, Yasuji
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10
For the development of nuclear fuel cycle, it is one of the most important tasks to improve reprocessing more economically and efficiently. Especially, to establish the Fast Breeder Reactor (FBR) cycle system for the future, it is strongly desirable to develop a new reprocessing which uses more compact equipments and produces less radioactive wastes compared to the present PUREX process. For this purpose, we have proposed a novel aqueous reprocessing system named ERIX Process to treat spent FBR-MOX fuels. This process consists of (1) Pd removal by selective adsorption using a specific anion exchanger; (2) electrolytic reduction for the valence adjustment of the major actinides including U, Pu, Np and some fission products (FP) such as Tc and Ru; (3) anion exchange separation for the recovery of U, Pu and Np using a new type of anion exchanger, AR-01; and (4) selective separation of long-lived minor actinides (MA = Am and Cm) by extraction chromatography. In this work, MA separation process was studied.
Mineo, Hideaki; Asakura, Toshihide; Hotoku, Shinobu; Ban, Yasutoshi; Morita, Yasuji
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.1250 - 1255, 2003/11
An advanced aqueous reprocessing process has been proposed for the next generation fuel cycle. Key technologies applied to the process are: removal of I-129, separation of Np and FP(Tc) separation by selective reduction of Np(VI) and high acid scrubbing of Tc within a single cycle process, MA separation by extraction chromatography and Cs/Sr separation. U separation just after dissolution was supposed to be effective to reduce the required capacity of the following extraction step. Among them Np reduction rate in TBP solution was measured, which was found to be lower than that in aqueous solution. Using an improved flow sheet spent fuel test, based on the Np reduction test, was carried out and about 90% of Np was separated before U and Pu partitioning step.
Kuwabara, Jun; Tolmachyov, S.; Noguchi, Hiroshi
Journal of Nuclear Science and Technology, 39(Suppl.3), p.556 - 559, 2002/11
no abstracts in English
; ;
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.
*; Arai, Tsuyoshi*; *
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.
Mineo, Hideaki; ; Takeshita, Isao; Nishizawa, Ichio; Sugikawa, Susumu; *
Nuclear Technology, 117(3), p.329 - 339, 1997/03
Times Cited Count：1 Percentile：84.66(Nuclear Science & Technology)no abstracts in English
Mineo, Hideaki; ;
JAERI-Tech 96-047, 42 Pages, 1996/10
no abstracts in English
systemAkatsu, Jokun; Kimura, Takaumi
Journal of Radioanalytical and Nuclear Chemistry, 140(1), p.195 - 203, 1990/00
Times Cited Count：9 Percentile：31.46(Chemistry, Analytical)no abstracts in English
; ;
Bunseki Kagaku, 23(7), p.764 - 768, 1974/07
no abstracts in English
; ;
Bunseki Kagaku, 22(4), p.437 - 443, 1973/04
no abstracts in English
