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.

Uranium and plutonium extraction by -dialkylamides using multistage mixer-settler extractors

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

Procedia Chemistry, 21, p.156 - 161, 2016/12

https://doi.org/10.1016/j.proche.2016.10.022

A continuous counter-current experiment was carried out to demonstrate the validity of a process using -dialkylamides for recovering U and Pu. This process consisted of two cycles, and the 1st cycle and the 2nd cycle employed -di(2-ethylhexyl)-2,2-dimethylpropanamide and -di(2-ethylhexyl)butanamide as extractants, respectively. The feed solution for the 1st cycle was 5.1 mol/dm (M) nitric acid containing 0.92 M U, 1.6 mM Pu, and 0.6 mM Np. The raffinate collected in the 1st cycle was used as the feed for the 2nd cycle. The ratios of U recovered in the U fraction and U-Pu fraction were 99.1% and 0.8%, respectively. The ratio of Pu recovered in the U-Pu fraction was 99.7%. The concentration ratio of U with respect to Pu in the U-Pu fraction was 9, and this indicated that Pu was not isolated. The decontamination factor of U with respect to Pu in the U fraction was obtained as 4.510. These results supported the validity of the proposed process.

Spent fuel test equipment for research and development activities at NUCEF-BECKY

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

Proceedings of 5th International Conference on Asian Nuclear Prospects 2016 (ANUP 2016) (CD-ROM), 2 Pages, 2016/10

The Japan Atomic Energy Agency (JAEA) conducts extensive research and development (R&D) activities toward the development of aqueous separation processes for spent nuclear fuel at the Nuclear Fuel Cycle Safety Engineering Research Facility-Back-end Fuel Cycle Key Element Research Facility (NUCEF-BECKY). BECKY is equipped with three hot cells, several glove boxes, and fume hoods. One of the hot cells, called the process cell, contains bench scale spent fuel test equipment used for U, Pu, and other spent fuel material. In this study, we describe an equipment used for aqueous separation processes in the process cell.

Supposed existence of Np in a genuine dissolver solution from the results of extraction simulation by PARC-L code

Asakura, Toshihide; Hotoku, Shinobu; Morita, Yasuji

Journal of Nuclear Science and Technology, 52(12), p.1552 - 1561, 2015/12

https://doi.org/10.1080/00223131.2015.1018361

With a genuine spent fuel soltuion (a dissolver solution), a laboratory-scale reprocessing experiment of an extraction-separation process was performed using mixer-settlers as extactors. In the experiment, n-butyraldehyde was utilized as a reducing reagent of NpO to NpO for the purpose to distinguish NpO from Np. From the Np concentration in the aqueous phase, Np would be extracted from the dissolver solution together with U and Pu. The scrutiny of Np behavior was performed utilizing 66 cases of calculation results by a Japan Atomic Energy Agency open extraction simulation code, the Programm for Advanced Extraction with Radiation Effect Calculation-Lightened version. From the scrutiny, the authors found that the calculation result with 60% of Np in the dissolver solution represented the best experimental extraction-separtion behavior of Np. Therefore, it was supposed that the dissolver solution contained sufficient proportion of Np to affect the extraction-separation behavior of Np.

Extraction separation of U and Pu by -di(2-)-2,2-dimethylpropanamide (DEHDMPA) and -di(2-)butanamide (DEHBA) using mixer-settlers in the presence of degradation products of DEHDMPA and DEHBA

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

Solvent Extraction Research and Development, Japan, 22(1), p.47 - 55, 2015/00

Two sets of continuous counter-current experiments using mixer-setters were performed for evaluating extraction properties of -dialkylamides toward U and Pu in the presence of their degradation products. The 1st cycle employed -di(2-)-2,2-dimethylpropanamide (DEHDMPA) for selective extraction of U(VI), and the 2nd cycle employed -di(2-)butanamide (DEHDBA) for co-extraction of U(VI) and Pu(IV). Degradation products were added to the organic phase of each cycle. Most of U was effectively extracted by DEHDMPA, and the ratio of U recovered to the U fraction was 99.57%. DEHDMPA barely extracted Pu, and the decontamination factor of U with respect to Pu in the U fraction was 1.1 10. The raffinate of the 1st cycle was used as the feed in the 2nd cycle, and the residual U and almost all Pu were effectively extracted by DEHBA. The degradation products had no detrimental effects on the two-phase separation and the operation of mixer-settlers.

Stabilization of uranium hexafluoride by hydrolysis method for decommissioning of safeguard laboratory facility

Inagawa, Jun; Hotoku, Shinobu; Oda, Tetsuzo; Aoyagi, Noboru; Magara, Masaaki

JAEA-Technology 2014-007, 48 Pages, 2014/03

JAEA-Technology-2014-007.pdf:5.76MB

In safeguard laboratory (SGL) facility of Nuclear Science Research Institute of JAEA, uranium hexafluoride (UF) of enriched uranium of various enrichment was used for research and development of a spectrometric method for the determination of the enrichment of uranium in April 1983 through March 1993. After completion of this R&D, the UF has been stored in SGL facility. It was decided that the UF is carried to out of the facility, because the SGL facility will be decommissioning until March 2015. To transport and store in safety after transportation, it is necessary that the UF should be converted to stable chemical form. Hydrolysis of UF to uranyl fluoride (UOF) and evaporation to solid state were selected for the stabilization method. The equipment for hydrolysis and evaporation was installed in the SGL facility. Stabilization was operated in this equipment, and all of the UF in the SGL facility was converted to UOF solid state in October 2012 through August 2013. In this report, results of examination and operation for stabilization of UF were reported.

Isotope dilution inductively coupled plasma mass spectrometry for determination of Sn content in spent nuclear fuel sample

Asai, Shiho; Toshimitsu, Masaaki; Hanzawa, Yukiko; Suzuki, Hideya; Shinohara, Nobuo; Inagawa, Jun; Okumura, Keisuke; Hotoku, Shinobu; Kimura, Takaumi; Suzuki, Kensuke*; et al.

Journal of Nuclear Science and Technology, 50(6), p.556 - 562, 2013/06

https://doi.org/10.1080/00223131.2013.785276

The Sn content in a spent nuclear fuel solution was determined by ICP-MS for its inventory estimation in high-level radioactive waste. An irradiated UO fuel was used as a sample to evaluate the reliability of the methodology. Prior to the measurement, Sn was separated from Te, which causes major isobaric interference in the determination of Sn content, along with highly radioactive coexisting elements using an anion-exchange column. The absence of counts attributed to Te in the Sn-containing effluent indicates that Te was completely removed. After washing, Sn retained on the column was readily eluted with 1 M HNO. The isotope ratios of Sn were successfully determined and showed good agreement with those obtained through ORIGEN2 calculations. The results reported in this paper are the first experimental values of Sn content in the spent nuclear fuel solution originating in spent nuclear fuel irradiated at a nuclear power plant in Japan.

Distribution of U(VI) and Pu(IV) by -di(2-ethylhexyl)butanamide in continuous counter-current extraction with mixer-settler extractor

Ban, Yasutoshi; Hotoku, Shinobu; Morita, Yasuji

Solvent Extraction and Ion Exchange, 30(2), p.142 - 155, 2012/02

https://doi.org/10.1080/07366299.2011.609388

The extraction properties of 1.5 mol/dm (M) -di-(2-ethyl-hexyl)butanamide (DEHBA) diluted with -dodecane toward U(VI) and Pu(IV) were studied using a single stage batch method, and the distribution ratios toward U(VI) and Pu(IV) were, respectively, obtained as follows: = 1.4[NO ][DEHBA] and = 0.11[NO ][DEHBA]. A continuous counter-current experiment with 1.5 M DEHBA as an extractant was performed using mixer-settler extractors. The ratios of U(VI) and Pu(IV) extracted by 1.5 M DEHBA in the U-Pu extraction step were more than 99.9%. The extracted Pu(IV) was back-extracted using 0.67 M nitric acid, and more than 97% of Pu(IV) in the feed was recovered in the Pu fraction. The present results indicated that DEHBA works as an extractant for mutual separation of U(VI) and Pu(IV) by adjusting the nitric acid concentration without using Pu(IV) reductants.

Determination of Se and Cs in spent nuclear fuel for inventory estimation of high-level radioactive wastes

Asai, Shiho; Hanzawa, Yukiko; Okumura, Keisuke; Shinohara, Nobuo; Inagawa, Jun; Hotoku, Shinobu; Suzuki, Kensuke*; Kaneko, Satoru*

Journal of Nuclear Science and Technology, 48(5), p.851 - 854, 2011/05

https://doi.org/10.3327/jnst.48.851

Technetium separation for future reprocessing

Asakura, Toshihide; Hotoku, Shinobu; Ban, Yasutoshi; Matsumura, Masakazu; Morita, Yasuji

Journal of Nuclear and Radiochemical Sciences, 6(3), p.271 - 274, 2005/12

https://doi.org/10.14494/jnrs2000.6.3_271

Tc extraction and separation experiments were performed basing on PUREX technique with using spent UO fuel with burn-up of 44 GWd/t. The experimental results were examined with performing calculations by a simulation code ESSCAR (Extraction System Simulation Code for Advanced Reprocessing). It was demonstrated that Tc can be almost quantitatively extracted from a dissolver solution and that Tc can also be almost quantitatively recovered by scrubbing. Further, it was clearly presented from the calculation results of ESSCAR that the extraction mechanism of Tc is dominated by the synergistic effect of Zr and U.