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Journal Articles

Actinides recovery from irradiated fuel for SmART cycle

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.

JAEA Reports

None

; *

JNC TN8200 2001-005, 54 Pages, 2001/08

JNC-TN8200-2001-005.pdf:5.85MB

None

JAEA Reports

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.

JAEA Reports

Summary of the dissolution experiments of the irradiated fast reactor fuels in CPF

; Koyama, Tomozo; Funasaka, Hideyuki

JNC TN8400 2000-016, 188 Pages, 2000/03

JNC-TN8400-2000-016.pdf:3.6MB

We summarized the conditions and results of all dissolution experiments (bench scale experiments (dissolution of sheared fuel pins) and beaker scale experiments (dissolution of a few sheared fuels pieces) of the irradiated fast reactor fuels, which were carried out in the Chemical Processing Facility (CPF). The fabrication and irradiation conditions of the dissolved fuels were also put in order.

JAEA Reports

Study about the dissolution behavior of the irradiated fast reactor fuels in CPF

; Koyama, Tomozo; Funasaka, Hideyuki

JNC TN8400 2000-014, 78 Pages, 2000/03

JNC-TN8400-2000-014.pdf:2.13MB

We investigated the factors which affected the dissolution of U and Pu to the nitric acid solution with the fragmentation model, which was based on the results of dissolution experiments for the irradiated fast reactor fuels in the Chemical Processing Facility(CPF). The equation that gave the fuel dissolution rate was estimated with the condition of fabrication (Pu ratio (Pu/(U+Pu))), irradiation (burn-up) and dissolution (nitric acid concentration, solution temperature and U+Pu concentration) by evaluating these effects quantitatively. We also investigated the effects of fuel volume ratio to the solution in the dissolver, burn-up and flouring ratio of the fuel on the f-value (the parameter which shows the diffusion and osmosis of nitric acid to the fuel) in the fragmentation model. It was confirmed that the fuel dissolution rate calculated with this equation had better agreement with the results of dissolution experiments for the irradiated fast reactor fuels in the CPF than that estimated with the surface area model. In addition, the efficiency of this equation was recognized for the dissolution of unirradiated U pellet and high Pu enriched MOX fuel. It was shown that the dissolution rate of the fuel slowed down at the condition of the high U-Pu concentration dissolution by the calculation of the dissolution behavior with this equation. The dissolution of the fuel can be improved by increasing the nitric acid concentration and temperature, but from the viewpoint of lowering the corrosion of the dissolver materials, it is desirable that the f-value is increased by optimizing the condition of shearing and stirring for the improvement of dissolution.

Journal Articles

Diffusion of metal fission products in ZrC$$_{1}$$$$_{.}$$$$_{0}$$

; Ikawa, Katsuichi

Journal of Nuclear Materials, 105, p.331 - 334, 1982/00

 Times Cited Count:28 Percentile:97.69(Materials Science, Multidisciplinary)

no abstracts in English

Oral presentation

Summary of the SmART cycle project of MA-MOX fuel which had extracted from Joyo spent fuel

Koizumi, Tsutomu

no journal, , 

no abstracts in English

Oral presentation

Minor actinoids (MA) separation and international cooperation with foreign countries

Koizumi, Tsutomu

no journal, , 

no abstracts in English

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