Conceptual design of iodine-sulfur process flowsheet with more than 50% thermal efficiency for hydrogen production

Kasahara, Seiji; Imai, Yoshiyuki; Suzuki, Koichi*; Iwatsuki, Jin; Terada, Atsuhiko; Yan, X.

Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.491 - 500, 2016/11

A conceptual design of a practical large scale plant of the thermochemical water splitting iodine-sulfur (IS) process flowsheet was carried out as a heat application of Japan Atomic Energy Agency's commercial Gas Turbine High Temperature Reactor Cogeneration (GTHTR300C) plant design. Innovative techniques proposed by JAEA were applied for improvement of hydrogen production thermal efficiency; flash concentration of HSO using waste heat from Bunsen reaction, prevention of HSO vaporization from a distillation column by introduction of HSO solution, and I condensation heat recovery by direct contact heat exchange in the HI distillation column. A simulation of material and heat balance showed hydrogen of about 31,900 Nm/h was produced by 170 MW heat from the GTHTR300C. A process thermal efficiency of 50.2% was achievable with incorporation of the innovative techniques and several high performance components expected in future R&D.

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

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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.

Reduction stripping of plutonium loaded on TBP with addition of nitrous acid

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Journal of Nuclear Science and Technology, 13(6), p.321 - 326, 1976/06

https://doi.org/10.3327/jnst.13.321

no abstracts in English

Development of U and Pu co-processing process in the Tokai Reprocessing Plant

Kudo, Atsunari; Yanagibashi, Futoshi; Hoshi, Takahiro; Tada, Kazuhito; Sato, Takehiko; Fujimoto, Ikuo; Obu, Tomoyuki

no journal, , 

no abstracts in English

Development of U and Pu co-processing process; Studies of U, Pu and Np co-recovery flow sheets with centrifugal contactors

Kurabayashi, Kazuaki; Kudo, Atsunari; Sato, Takehiko; Tada, Kazuhito*; Obu, Tomoyuki

no journal, , 

We are developing Np recovery with U and Pu in co-processing process (U and Pu co-recovery) because Np is one of minor actinides and a long-lived radionuclide. It is reported the flow sheet of U, Pu and Np co-recovery we set by MIXSET-X.