Conceptual Design Study on Advanced Aqueous Reprocessing System for Fast Reactor Fuel Cycle

Takata, Takeshi; Koma, Yoshikazu; Sato, Kōji; Kamiya, M.; Shibata, Atsuhiro; Nomura, Kazunori; Ogino, Hideki; Koyama, Tomozo; Aose, Shinichi

doi:10.1080/18811248.2004.9715489

Cited by 28 publications

(22 citation statements)

References 2 publications

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“…In response to these problems, an engineering-scale continuous crystallizer has been developed and successfully operated for the crystallization of UNH. 4,13,22) It has also been demonstrated using a beaker-scale experiment that Pu(IV) can be successfully oxidized to Pu(VI) by heating the solution. 7) Although other problems are still under consideration, much knowledge regarding solution chemistry and rich experience accumulated during the development of the PUREX processes are available for the development of the U-Pu co-crystallization system.…”

Section: Discussionmentioning

confidence: 99%

“…3) Recently, in Japan, crystallization has been considered as a principal separation method in the NEXT (New EXtraction System for TRU Recovery) process, 4) which is designed for fast reactor fuel. In the NEXT process, approximately 70% of the uranium is recovered by crystallization prior to solvent extraction in order to reduce the mass flow of the solution to the PUREX extraction processes.…”

Section: Introductionmentioning

confidence: 99%

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Flowsheet Study of U-Pu Co-Crystallization Reprocessing System

Homma

Ishii

Kikuchi

et al. 2008

J. Nucl. Sci. Technol.

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A U-Pu co-crystallization reprocessing system is proposed for light water reactor fuels and its flowsheet study is carried out. This reprocessing system is based on experimental evidence indicating that Pu(VI) in a nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate is not present in the solution. The system consists of five steps: dissolution of spent fuel, Pu oxidation, U-Pu co-crystallization, re-dissolution of the crystals, and U re-crystallization. The proposed system does not require the use of organic solvents, resulting in a relative increase in safety and cost-effectiveness. The system requires recycling of the mother liquor from the U-Pu co-crystallization step to recover almost the entire amount of U and Pu at this step. The appropriate recycling ratio is determined, such that satisfactory decontamination is achieved. A consistent ratio of Pu to U in the mother liquor from the U re-crystallization is maintained by regulating the temperature, suggesting that the quality of the liquor, which can be a source of mixed oxide fuels, can be controlled despite differences in the composition of the spent fuel. The size of a plant utilizing the proposed system is estimated to be about 30% less than that of the PUREX system.

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Section: Discussionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Flowsheet Study of U-Pu Co-Crystallization Reprocessing System

Homma

Ishii

Kikuchi

et al. 2008

J. Nucl. Sci. Technol.

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“…where W 0 , f , r(t), S 0 , and F(f) are the initial amount of fuel (mol), the dissolved ratio (-), the instantaneous dissolution rate (mol cm 72 min 71 ), the initial geometrical surface area of the cross section of sheared fuel pieces (cm 2 ), and the transitional function of the surface area (-), respectively. Nemoto et al [5] reported that the instantaneous dissolution rate and transitional function of the surface area of irradiated MOX fuel (sheared with 30 mm in length) were described as: [5].…”

Section: Surface-area Modelmentioning

confidence: 99%

“…The dissolution rate of MOX fuel is depressed upon increasing the Pu content [1]. On the other hand, a modified PUREX process has been suggested for FR fuel reprocessing, which includes a crystallization process for rough uranium recovery prior to the simplified extraction process [2]. For the required uranium recovery to be achieved in the crystallization process, the dissolver solution has to be prepared with a high HM concentration of more than 500 g dm 73 [2].…”

Section: Introductionmentioning

confidence: 99%

“…On the other hand, a modified PUREX process has been suggested for FR fuel reprocessing, which includes a crystallization process for rough uranium recovery prior to the simplified extraction process [2]. For the required uranium recovery to be achieved in the crystallization process, the dissolver solution has to be prepared with a high HM concentration of more than 500 g dm 73 [2]. Under such conditions, fuel dissolution will be substantially slower because the highly concentrated condition requires a larger amount of fuel to be treated and a smaller volume of nitric acid.…”

Section: Introductionmentioning

confidence: 99%

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Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

Ikeuchi

Sano

Shibata

et al. 2013

Journal of Nuclear Science and Technology

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An efficient dissolution process was established for future reprocessing in which mixed-oxide (MOX) fuels with high plutonium contents and dissolver solution with high heavy-metal (HM) concentrations (more than 500 g dm 73 ) will be treated. This dissolution process involves short stroke shearing of fuels (*10 mm in length). The dissolution kinetics of irradiated MOX fuels and the effects of the Pu content, HM concentration, and fuel form on the dissolution rate were investigated. Irradiated fuel was found to dissolve as 10 2 -10 3 times fast as non-irradiated fuel, but the rate decreased with increasing Pu content. Kinetic analysis based on the fragmentation model, which considers the penetration and diffusion of nitric acid through fuel matrices prior to chemical reaction, indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid (L/S ratio) but also by the exposed surface area per unit mole of nitric acid (A/m ratio). The penetration rate of nitric acid is expected to be decreased at high HM concentrations by a reduction in the L/S ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the A/m ratio.

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Uranyl nitrate crystallizer performance with changing solution level

Nadezhdin

Gozhimov

Goryunov

et al. 2019

Heliyon

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This manuscript described a dynamic simulation of uranyl nitrate crystallization in a linear crystallizer. Furthermore, a mathematical model of the crystallizer supply system was developed with a related control algorithm; the model contained two piston feeders. The results showed the crystallization process's sensitivity to the solution level in the crystallizer. An expression for calculating the performance of the crystallizer was proposed. That expression allowed us to understand that the accuracy of the liquid phase level (to avoid the crystallizer's performance decreasing by more than 5%) should be in the range of ±4% of the crystallization section height. For this, we developed a control algorithm for the supply system to support operability. This algorithm allowed us to maintain a specified level of mother solution in the crystallization area and provided an asynchronous operation mode for the piston batchers. Furthermore, this paper described how the developed mathematical model and the proposed control system, i.e., the envisaged recommendations, can be used to optimize the process during the design and adjustment of equipment.

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Conceptual Design Study on Advanced Aqueous Reprocessing System for Fast Reactor Fuel Cycle

Cited by 28 publications

References 2 publications

Flowsheet Study of U-Pu Co-Crystallization Reprocessing System

Flowsheet Study of U-Pu Co-Crystallization Reprocessing System

Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

Uranyl nitrate crystallizer performance with changing solution level

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