A Demonstration test to separate minor actinides in high-level liquid waste by -hexaoctyl nitrilotriacetamide (HONTA) using mixer-settler extractors in a hot cell

Ban, Yasutoshi; Suzuki, Hideya*; Hotoku, Shinobu; Tsubata, Yasuhiro

Solvent Extraction Research and Development, Japan, 31(1), p.1 - 11, 2024/00

https://doi.org/10.15261/serdj.31.1

A demonstration test was performed to separate minor actinides (MA; Am and Cm) by -hexaoctyl nitrilotriacetamide (HONTA) as an extractant using mixer-settler extractors installed in a hot cell. A high-level liquid waste containing MA, and rare earths (RE; Y, La, Nd, and Eu) was used as the feed. HONTA diluted to 0.05 mol/dm in -dodecane was fed as the organic phase, and a part of the organic phase was reused without solvent regeneration. HONTA effectively extracted MA, whereas RE were less extractable. Consequently, the Y, La, Nd, and Eu ratios distributed to a RE fraction were 99.9%, 99.2%, 61.8%, and 81.4%, respectively. The Am and Cm ratios distributed to an MA fraction were 86.8% and 74.7%, respectively, and a substantial amount of MA (0.12 g) was recovered in the MA fraction by the end of the cumulative duration of 40 h.

Minor actinides separation by -hexaoctyl nitrilotriacetamide (HONTA) using mixer-settler extractors in a hot cell

Ban, Yasutoshi; Suzuki, Hideya*; Hotoku, Shinobu; Tsutsui, Nao; Tsubata, Yasuhiro; Matsumura, Tatsuro

Solvent Extraction and Ion Exchange, 37(7), p.489 - 499, 2019/11

https://doi.org/10.1080/07366299.2019.1693486

A continuous counter-current experiment to separate minor actinides (MAs: Am and Cm) was performed with -hexaochyl nitrilotriacetamide (HONTA) as an extractant. Nitric acid of 0.08 M (mol/dm) containing MAs and rare earths (REs) recovered from high-level waste was used as the Feed, and the experiment was conducted for 14 h. The ratios of Am and Cm recovered into the MA fraction measured 94.9% and 78.9%, respectively. HONTA hardly extracted Y, La, and Eu in the Feed (99.9% for Y, 99.9% for La, and 96.7% for Eu), most of which were distributed to the RE fraction. A portion of Nd was extracted by HONTA, and consequently the ratio of Nd in the RE fraction was 83.5%. The concentrations of MAs and some REs in each stage were calculated using a simulation code, and the results are consistent with the experimental values. This code indicates that the ratios of MAs in the MA fraction and REs in the RE fraction could be 99% by optimizing separation conditions.

Status of fuel transmutation programmes in Japan and France; Lessons drawn from results

Arai, Yasuo; Pillon, S.*

Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles (CD-ROM), 9 Pages, 2004/06

no abstracts in English

On the electronic structure of CmF(=1-4) by all-electron Dirac-Hartree-Fock calculations

Mochizuki, Yuji*; Tatewaki, Hiroshi*

Journal of Chemical Physics, 118(20), p.9201 - 9207, 2003/05

https://doi.org/10.1063/1.1568075

no abstracts in English

Lattice parameter expansion by self-irradiation damage of Cm-Pu oxide and mononitride

Takano, Masahide; Ito, Akinori; Akabori, Mitsuo; Ogawa, Toru; Numata, Masami; Kizaki, Minoru

Journal of Nuclear Science and Technology, 39(Suppl.3), p.842 - 845, 2002/11

http://www.tandfonline.com/doi/abs/10.1080/00223131.2002.10875600

no abstracts in English

On the electronic structure of Cm(HO)(n=1, 2, 4, 6) by all-electron Dirak-Hartree-Fock calculations

Mochizuki, Yuji*; Tatewaki, Hiroshi*

Journal of Chemical Physics, 116(20), p.8838 - 8842, 2002/05

https://doi.org/10.1063/1.1473803

no abstracts in English

Luminescence study on solvation of americium(III), curium(III) and several lanthanide(III) ions in nonaqueous and binary mixed solvents

Kimura, Takaumi; Nagaishi, Ryuji; Kato, Yoshiharu; Yoshida, Zenko

Radiochimica Acta, 89(3), p.125 - 130, 2001/05

https://doi.org/10.1524/ract.2001.89.3.125

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.

Scoping calculation of nuclides migration in engineering barrier system for effect of volume expansion due to overpack corrosion and intrusion of the buffer material

; ; Ishiguro, Katsuhiko; Nakajima, Kunihiko*;

JNC TN8400 99-087, 41 Pages, 1999/11

JNC-TN8400-99-087.pdf:7.99MB

Corrosion of the carbon steel overpack leads to a volume expansion since the specific gravity of corrosion products is smaller than carbon steel. The buffer material is compressed due to the corrosive swelling, reducing its thickness and porosity. On the other hand, Buffer material may be extruded into fractures of the surrounding rock and this may lead to a deterioration of the planned functions of the buffer, including retardation of nuclides migration and colloid filtration. In this study, the sensitivity analyses for the effect of volume expansion and intrusion of the buffer material on nuclide migration in the engineering barrier system are carried out. The sensitivity analyses were performed on the decrease in the thickness of the buffer material in the radial direction caused by the corrosive swelling, and the change in the porosity and dry density of the buffer caused by both compaction due to corrosive swelling and intrusion of buffer material. As results, it was found the maximum release rates of relatively shorter half-life nuclides from the outside of the buffer material decreased for taking into account of a volume expansion due to overpack corrosion. On the other hand, the maximum release rates increased when the intrusion of buffer material was also taking into account. It was, however, the maximum release rates of longer half-life nuclides, such as Cs-137 and Np-237, were insensitive to the change of buffer material thickness, and porosity and dry density of buffer.

Thermodynamic Date for Predicting Concentrations of Pu(III), Am(III), and Cm(III) in Geologic Environments

Rai, D.*; Rao, L.*; Weger, H. T.*; GREGORY R.CHOPPI*; Yui, Mikazu

JNC TN8400 99-010, 95 Pages, 1999/01

JNC-TN8400-99-010.pdf:3.88MB

This report provides thermodynamic data for predicting concentrations of Pu(III), Am(III), and Cm(III) in geologic environments, and contributes to an integration of the JNC chemical thermodynamic database, JNC-TDB (previously PNC-TDB), for the performance analysis of geological isolation system for high-level radioactive wastes. Thermodynamic data for the formation of complexes or compounds with hydroxide, chloride, fluoride, carbonate, nitrate, sulfate and phosphate are discussed in this report. Where data for specific actinide(III) species are lacking, the data were selected based on chemical analogy to other trivalent actinides. In this study, the Pitzer ion-interaction model is mainly used to extrapolate thermodynamic constants to zero ionic strength at 25C.