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Development of hybrid reprocessing technology based on solvent extraction and pyro-chemical electrolysis
Progress in Nuclear Energy 53 (2011) 940e943
Contents lists available at ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Development of hybrid reprocessing technology based on solvent extraction and pyro-chemical electrolysis Hisao Ohmura a, *, Koji Mizuguchi a, Shohei Kanamura a, Tetsuo Ohsato a, Reiko Fujita a, Takashi Omori b, Kazuhiro Utsunomiya b a b
Power & Industrial Research & Development Center, Toshiba Corporation, 4-1, Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862, Japan Chemical System Design & Engineering Department, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama 235-8523, Japan
a r t i c l e i n f o
a b s t r a c t
Article history: Received 28 October 2010 Received in revised form 27 May 2011 Accepted 30 May 2011
Toshiba has been proposing a new fuel cycle concept for the transition period from Light Water Reactors (LWRs) to Fast Reactors (FRs). This concept involves a more valuable process for LWR spent-fuel reprocessing than the conventional process and improved proliferation resistance. We have been developing a new technology, the Toshiba Hybrid Reprocessing Process, based on solvent extraction and pyro-chemical electrolysis, for spent fuel reprocessing for the transition period from LWRs to FRs. The Toshiba Hybrid Reprocessing Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high-purity uranium (U) and the pyro-chemical process in molten salts for recovery of impure plutonium with minor actinides (Pu þ MA). High-purity U is used for LWR fuel, and impure Pu þ MA is used for metallic FR fuel. Valence control by the electrolysis and solvent extraction tests using LWR spent fuel and oxalate precipitation tests were carried out to confirm the feasibility of the Toshiba Hybrid Reprocessing Process. A consecutive processing equipment for the solvent extraction process and a bench-scale apparatus for the pyro-chemical process were manufactured. The consecutive processing equipment consists of a flow type electrolytic cell and a centrifugal extractor. The test revealed that U of 99.99% of purity was recovered. The bench-scale apparatus consists of a reactor for oxalate precipitation, a solideliquid separator in which nitric acid with fission products and precipitation are separated, and a drying equipment in which the precipitation is dry. Precipitation test with neodymium (Nd) which is simulated as Pu þ MA in nitric acid was carried out. It was confirmed that precipitation ratio of Nd was more than 99.9% and that moisture ratio of the precipitation was less than 10%. The results suggested that U recovery of LWR spent fuel was 99.99% with the consecutive processing equipment and Pu þ MA recovery was more than 99.9% with the bench-scale apparatus. The Toshiba Hybrid Reprocessing Process could recover high-purity U used for LWR fuel, and impure Pu þ MA used for metallic FR fuel. Ó 2011 Elsevier Ltd. All rights reserved.
Keywords: Hybrid reprocessing Transition Valence control Oxalate precipitation
1. Introduction For next-generation Fast Reactors (FRs), it will be essential to realize a nuclear fuel cycle in which spent fuel is reprocessed and reused. As mentioned in the Framework for Nuclear Energy Policy (October 2005), if Fast Reactors come online after 2050, they will coexist with Light Water Reactors (LWRs) until at least around 2100. Toshiba has developed a reprocessing process based on hybrid reprocessing with advanced solvent extraction and pyrochemical electrolysis that can be applied to the nuclear fuel cycle from the time of introduction of Fast Reactors through to their
* Corresponding author. Tel.: þ81 44 288 8158; fax: þ81 44 270 1807. E-mail address: [email protected] (H. Ohmura). 0149-1970/$ e see front matter Ó 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2011.05.033
complete adoption, and aims to use this as the basis of a nextgeneration nuclear fuel cycle (Fig. 1) (Mizuguchi et al., 2008, 2010a). Most several other concepts for reprocessing technology in transition period from LWRs and FRs add several new process to conventional reprocessing based on the PUREX process. The conventional reprocessing based on the PUREX solvent extraction method, is suitable for the recovery of uranium with a high decontamination factor. However, this process involves a high risk of nuclear proliferation because the recovery of plutonium with a low decontamination factor is difficult in this process. This process involves many high level wastes generates because minor actinide become high level wastes. Toshiba has been developing a new technology, the Toshiba Hybrid Process, for spent-fuel reprocessing based on solvent extraction and pyro-chemical electrolysis, suitable for the
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Fig. 3. The method for the advanced extraction.
2. Uranium extraction with electrolytic reduction Fig. 1. Toshiba’s concept of next-generation nuclear fuel cycle.
transition period from LWRs to FRs (Fig. 2) reprocessing because it can produces fuel for FR by reprocessing spent fuel for LWR (Mizuguchi et al., 2008, 2010a). The Hybrid Process combines the solvent extraction process of LWR spent fuel in nitric acid with the recovery of high-purity uranium and the pyro-chemical process in molten salts for the recovery of impure plutonium with minor actinides. High-purity uranium is used for LWR fuel, and impure plutonium with minor actinides is used for metallic FR fuel. The pyro-chemical process for the FR spent fuel recycling system is based on research on an electrorefining process in molten salts conducted at Toshiba since 1988(Koyama et al., 1995; Kobayashi et al., 1997; Koyama et al., 1997; Iizuka et al., 1997; Mizuguchi et al., 2009; Fujita et al., 2008). The new Toshiba Hybrid Process can reduce the amount of high level waste and secondary waste from spent-fuel reprocessing plants, and attain improvement in nuclear proliferation because impure plutonium is recovered with minor actinides for metallic FR fuel. As a first step, an advanced solvent extraction process have been developed in the Toshiba Hybrid Process (Fig. 2)(Mizuguchi et al., 2008, 2010a). The solvent extraction test with electrolytic reduction tests using actual LWR spent fuel were carried out to investigate the uranium recovery purity (target impurity of uranium is less than 1.0 106). As a second step, an Advanced Aqua-Pyro Process have been developed. In this process, actinides and rare earth elements are recovered from an aqueous solution in the extraction process by an oxalic acid precipitation method and are applied to the pyro-chemical electrolysis process. Oxalate acid precipitation tests were carried out in order to evaluate the recovery yield of Pu þ MA (target yield 99.9%).
Fig. 2. Toshiba Hybrid Process based on advanced solvent extraction and pyroelectrolysis.
An advanced solvent extraction process in the Hybrid Reprocessing is shown in Fig. 2. The key technology in this process is the method for the uranium extraction with electrolytic reduction shown in Fig. 3. This technology is based on the PUREX process using 30% tributyl phosphate (TBP) with 70% n-dodecane(n-DD) as an extractant. The method has a single extraction cycle consisting of extraction and electrolytic reduction. This method can extract only uranium by controlling the valence of plutonium ions to the trivalent state by electrolysis. Plutonium is kept in an aqueous solution with minor actinides and can be recovered with minor actinides in the pyro-chemical electrolysis. The hybrid process has high nuclear proliferation resistance because the recovery of plutonium with a high decontamination factor is difficult. Fig. 4 shows the scheme of the uranium extraction with electrolytic reduction. First, uranyl ions (UO2þ 2 ) are reduced to tetravalent uranous (U4þ) ions by electrolysis, and then these uranous ions act as a reductant to reduce tetravalent plutonium ions (Pu4þ) to trivalent ions (Pu3þ). Hexavalent plutonium ions are reduced to trivalent plutonium by tetravalent uranous ions. In the process, the reductant uranous ions are re-oxidized to uranyl ions (Mizuguchi et al., 2008, 2010a,b). Gram-scale actual spent fuel examinations had been carried out, and basic performance of the uranium extraction with electrolytic reduction was confirmed. In view of the gram-scale examinations results, continuous performance tests were carried out, and the impurity of recovered uranium was evaluated. A consecutive processing equipment shown in Fig. 5 was manufactured for the solvent extraction process. Continuous actual spent fuel examinations were carried out using the consecutive processing equipment. The equipment consists of a flow-typed electronic cell, a centrifugal contractor, and four pumps. The cell consists of an anode, a cathode, and a diaphragm made of inorganic material. This cell consists of two compartments.
Fig. 4. The scheme of the uranium extraction with electro-reduction.
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Fig. 6. Oxalate precipitation test using simulated spent fuel.
3. Advanced aqua-pyro process
Fig. 5. Consecutive processing equipment for the solvent extraction process.
The experimental conditions are shown in Table 1. These experiments were carried out at 0.5 mA/cm2. The burnup of actual spent fuel is about 65.9 GWd/tU. It was used as 0.5M-UO2þ 2 in 1M HNO3 solution. The results are given in Table 2. As result, the impurity of recovered uranium was decreased from 1.2 102 to about 1/70 into 1.7 104. The result suggests that the impurity of recovered uranium achieved less than the target value of 1.0 106 in a few stages. It was confirmed that uranium could be recovered with high decontamination factors.
At first an aqua-pyro process was developed (Akai and Fujita, 1995; Akai and Fujita, 1996; Akai and Fujita, 1997; Akai and Fujita, 1998a; Fujita and Akai, 1998b) in which minor actinides (MAs) are recovered from high-level radioactive liquid waste (HLLW) generated by operation of the Rokkasho Reprocessing Plant. This process is an electrolysis technology cultivated from pyro-chemical reprocessing. Next, electrolysis technology based on this aqua-pyro process was applied to an active reprocessing process, to realize a reprocessing process that makes it difficult to separate pure plutonium, while reducing the amount of HLLW. This process introduces reprocessing for LWR fuel that ensures a high level of support for nuclear non-proliferation. Subsequently, pyrochemical reprocessing technology based on electrolysis will be applied to reprocessing of FR fuel. This approach will allow progression of the core technology from the coexistence of FRs and LWRs through to complete adoption of FRs. As a result, Toshiba will be able to offer a fuel cycle suited to both LWRs and FRs in turn. An Advanced Aqua-Pyro Process (Mizuguchi et al., 2008, 2010a,b; Ohmura et al., 2010) has been developed by applying the aqua-pyro process in order to connect the advanced uranium extraction process with the pyroelectrolysis process. In this process, actinides and rare earth elements are recovered from the aqueous solution from the extraction process by the oxalic acid precipitation method and are applied to the pyro-chemical process. Basic oxalate precipitation tests were carried out on simulated spent fuel solution. Fig. 6 shows the test setup. The experimental conditions of the basic tests are shown in Table 3. Precipitation test with neodymium which is simulated as Pu þ MA in nitric acid was carried out because chemical nature of neodymium is similar to TRUs (Akai and Fujita, 1995; Akai and Fujita, 1996; Akai and Fujita, 1997; Akai and Fujita, 1998a; Fujita and Akai, 1998b). Uranium and neodymium were precipitated as uranyl oxalate and neodymium oxalate by the addition of oxalic acid. In the experiments at amount of oxalate ¼ 6 or more, the precipitation ratios of neodymium achieved the target value of 99% (Fig. 7). According to the reference about precipitation tests of rare earth elements by oxalic acid, solubility of neodymium oxalate is comparable as that of other rare earth oxalates. And according to
Table 2 Experimental results.
Table 3 Experimental conditions of basic test.
Table 1 Experimental conditions. 2þ
Concentration of SF
0.5M UO2
Current Flux of SF through electronic cell Flux of SF through bypass line Flux of 30% TBP with 70% n-DD
10A 10 ml/min 90 ml/min 100 ml/min
in 1M-HNO3aq
After supplying actual spent fuel to the cell and bypass line of the cell, a DC electric potential was applied to electrodes to cause 4þ 4þ in the actual spent the reaction of uranium from UO2þ 2 to U . Pu fuel reduce to Pu3þ by U4þ at a point which actual spent fuel through the cell and the actual spent fuel through the bypass line 4þ at the mix. U4þ in the actual spent fuel oxidized to UO2þ 2 by Pu mixing point at the same time. The actual spent fuel after the reaction sent to the centrifugal contractor. And then supplying 30% TBP with 70% n-DD to the centrifugal contractor, much uranium in the actual spent fuel was extracted by the 30% TBP with 70% n-DD, separated into the actual spent fuel and 30% TBP with 70% n-DD. The decontamination factors DFM were calculated according to the following Eq. (1), where C is the concentration of elements.
DFM ¼
CMfeed ,CUTBP CUfeed ,CMTBP
ðM ¼ Pu; Np; Am; CmÞ
(1)
DF Pu MA
682 Np Am Cm
37 953 74
Temperature Amount of oxalate Reaction time Concentration of U Concentration of Nd Concentration of HNO3aq
70 C 1e20 times the stoichiometric 0.5 h 0.1 M 0.033 M 0.1 N
H. Ohmura et al. / Progress in Nuclear Energy 53 (2011) 940e943
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Table 4 Experimental results of the bench-scale tests. Precipitation ratio Moisture ratio of the precipitation
99.9% or more 4.3%
in argon atmosphere and its moisture is removed before sending to molten salt It is enough to dry to the grade which can be treated as a solid in the advanced aqua-pyro process. Since neodymium could be recovered 99%, it is thought that it is recoverable 99% also about actinides or rare earth elements in actual spent fuel. 4. Conclusion
Fig. 7. Oxalate precipitation test using simulated spent fuel.
the reference, solubility of americium, curium, and plutonium oxalate are about 10 mg/L( ¼ 4 105 M), about 10 mg/L( ¼ 4 105 M), and about 100 mg/L( ¼ 4 104 M) respectively (Rudisill, 1996). Americium, curium, and plutonium contain in actual spent fuel by the same concentration as our experiments, 99% or more of these actinides would be recover in the same conditions. Since neodymium could be recovered 99%, it is thought that it is recoverable 99% also about actinides and rare earth elements in actual spent fuel. It was confirmed to recover 99% of actinides and rare earth elements from simulated spent fuel solution. A bench-scale apparatus shown in Fig. 8 was manufactured for the pyro-chemical process. The bench-scale apparatus consists of a reactor for oxalate precipitation, a solideliquid separator in which nitric acid with fission products and precipitation are separated, and a drying equipment in which the precipitation is dry. It was confirmed that precipitation ratio of neodymium was more than 99.9% and that moisture ratio of the precipitation was less than 10%(Table 4). After this process, the precipitation is further heated
Fig. 8. Bench-scale apparatus for the pyro-chemical process.
Toshiba has proposed a new fuel cycle concept for the transition period from LWRs to FRs. We have been developing a new technology, the Toshiba Hybrid Process, based on solvent extraction and a pyro-chemical process for spent-fuel reprocessing in this transition period. The Toshiba Hybrid Reprocessing Process could recover high-purity U used for LWR fuel, and impure Pu þ MA used for metallic FR fuel. References Akai, Y., Fujita, R., 1995. Development of transuranium element recovery from high level radioactive liquid waste. J. Nucl. Sci. Technol. 32 (10), 807e809. Akai, Y., Fujita, R., 1996. Development of transuranium element recovery from highlevel radioactive liquid waste - conversion of TRU oxalates to chlorides. J. Nucl. Sci. Technol. 33 (10). Akai, Y., Fujita, R., 1997. Development of Transuranium element recovery from highlevel radioactive liquid waste. In: Proc. Int. Conf. on Future Nuclear Systems, Global’97, Yokohama, Japan, Oct. 5-10, 1997. Fujita, R., Akai, Y., 1998a. Development of transuranium elements recovery from high-level radioactive liquid waste. J. Alloys Compounds, 271e273. Fujita, R., Akai, Y., 1998b. Separation of transuranium elements (TRUs) from rare earth elements by electrorefining in molten salts. In: Proc. 49th Annu. Meet. Int. Soc. Electrochemistry, Kokura, Japan, Sept. 15e17. Fujita, R., Mizuguchi, K., Fuse, K., Saso, M., Utsunomiya, K., Abe, K., 2008. Advanced hybrid process with solvent Extraction and pyro-chemical process of spent fuel reprocessing for LWR to FBR. In: 16Th Pac. Basin Nucl. Conf. (16PBNC), Aomori, Japan, Oct. 13-18. Iizuka, M., Koyama, T., Kondo, N., Fujita, R., Tanaka, H., 1997. Actinides recovery from molten salt/liquid metal system by electrochemical methods. J. Nucl. Mater. 247, 183e190. Kobayashi, T., Fujita, R., Nakamura, H., Koyama, T., 1997. Evaluation of cadmium pool potential in an electrorefiner with ceramic partition for spent metallic fuel. J. Nucl. Sci. Technol. 34 (1), 50e57. Koyama, T., Fujita, R., Iizuka, M., Sumida, Y., 1995. Pyrometallurgical reprocessing of fast reactor metallic fuel, -development of a new electrorefiner with a ceramic partition. Nucl. Tech. 110 (6), 357e368. Koyama, T., Iizuka, M., Shoji, Y., Fujita, R., Tanaka, H., Kobayashi, T., Tokiwai, M., 1997. An experimental study of molten salt electrorefining of uranium using solid Iron cathode and liquid cadmium cathode for development of pyrometallurgical reprocessing. J. Nucl. Sci. Technol. 34 (3), 384e393. Mizuguchi K., Fuse K., Nakamura H., Fujita R., Omori T., Utsunomiya K. "Development of Hybrid Reprocessing Technology with Solvent Extraction and Pyrochemical Electrolysis," Proc. Nihon-Genshiryoku-Gakkai, Kochi, Japan, (Sep. 4e6e2008), [in Japanese]. Mizuguchi, K., Fuse, K., Kanamura, S., Fujita, R., Omori, T., Utsunomiya, K., 2009. Development of hybrid reprocessing technology based on solvent extraction and pyrochemical electrolysis. In: Proc. International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities FR09, Kyoto, Japan, Dec. 9. Mizuguchi K., Ohmura H., Kanamura S., Fuse K., Nakamura H., Omori T., Utsunomiya K., Fujita R. "Development of Hybrid Reprocessing Technology (II)," Proc. NihonGenshiryoku-Gakkai, Ibaraki, Japan, (Mar. 26e28e2010a), [in Japanese]. Mizuguchi K., Ohmura H., T. Omori, K. Utsunomiya " Development of Hybrid Reprocessing Technology (I)," Proc. Nihon-Genshiryoku-Gakkai, Ibaraki, Japan, (Mar. 26e28e2010b), [in Japanese]. Ohmura H., Kanamura S., Fuse K., Mizuguchi K., Nakamura H., Omori T., Utsunomiya k., Fujita R. "Development of Hybrid Reprocessing Technology (III)," Proc. Nihon-Genshiryoku-Gakkai, Ibaraki, Japan, (Mar. 26e28e2010), [in Japanese] Rudisill, T.S., 1996. Pretreatment of Americium/Curium Solutions for Vitrification(U). Westinghouse Savannah River Co., Aiken, SC. WSRC-TR-96e0074.
Contents lists available at ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Development of hybrid reprocessing technology based on solvent extraction and pyro-chemical electrolysis Hisao Ohmura a, *, Koji Mizuguchi a, Shohei Kanamura a, Tetsuo Ohsato a, Reiko Fujita a, Takashi Omori b, Kazuhiro Utsunomiya b a b
Power & Industrial Research & Development Center, Toshiba Corporation, 4-1, Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862, Japan Chemical System Design & Engineering Department, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama 235-8523, Japan
a r t i c l e i n f o
a b s t r a c t
Article history: Received 28 October 2010 Received in revised form 27 May 2011 Accepted 30 May 2011
Toshiba has been proposing a new fuel cycle concept for the transition period from Light Water Reactors (LWRs) to Fast Reactors (FRs). This concept involves a more valuable process for LWR spent-fuel reprocessing than the conventional process and improved proliferation resistance. We have been developing a new technology, the Toshiba Hybrid Reprocessing Process, based on solvent extraction and pyro-chemical electrolysis, for spent fuel reprocessing for the transition period from LWRs to FRs. The Toshiba Hybrid Reprocessing Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high-purity uranium (U) and the pyro-chemical process in molten salts for recovery of impure plutonium with minor actinides (Pu þ MA). High-purity U is used for LWR fuel, and impure Pu þ MA is used for metallic FR fuel. Valence control by the electrolysis and solvent extraction tests using LWR spent fuel and oxalate precipitation tests were carried out to confirm the feasibility of the Toshiba Hybrid Reprocessing Process. A consecutive processing equipment for the solvent extraction process and a bench-scale apparatus for the pyro-chemical process were manufactured. The consecutive processing equipment consists of a flow type electrolytic cell and a centrifugal extractor. The test revealed that U of 99.99% of purity was recovered. The bench-scale apparatus consists of a reactor for oxalate precipitation, a solideliquid separator in which nitric acid with fission products and precipitation are separated, and a drying equipment in which the precipitation is dry. Precipitation test with neodymium (Nd) which is simulated as Pu þ MA in nitric acid was carried out. It was confirmed that precipitation ratio of Nd was more than 99.9% and that moisture ratio of the precipitation was less than 10%. The results suggested that U recovery of LWR spent fuel was 99.99% with the consecutive processing equipment and Pu þ MA recovery was more than 99.9% with the bench-scale apparatus. The Toshiba Hybrid Reprocessing Process could recover high-purity U used for LWR fuel, and impure Pu þ MA used for metallic FR fuel. Ó 2011 Elsevier Ltd. All rights reserved.
Keywords: Hybrid reprocessing Transition Valence control Oxalate precipitation
1. Introduction For next-generation Fast Reactors (FRs), it will be essential to realize a nuclear fuel cycle in which spent fuel is reprocessed and reused. As mentioned in the Framework for Nuclear Energy Policy (October 2005), if Fast Reactors come online after 2050, they will coexist with Light Water Reactors (LWRs) until at least around 2100. Toshiba has developed a reprocessing process based on hybrid reprocessing with advanced solvent extraction and pyrochemical electrolysis that can be applied to the nuclear fuel cycle from the time of introduction of Fast Reactors through to their
* Corresponding author. Tel.: þ81 44 288 8158; fax: þ81 44 270 1807. E-mail address: [email protected] (H. Ohmura). 0149-1970/$ e see front matter Ó 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2011.05.033
complete adoption, and aims to use this as the basis of a nextgeneration nuclear fuel cycle (Fig. 1) (Mizuguchi et al., 2008, 2010a). Most several other concepts for reprocessing technology in transition period from LWRs and FRs add several new process to conventional reprocessing based on the PUREX process. The conventional reprocessing based on the PUREX solvent extraction method, is suitable for the recovery of uranium with a high decontamination factor. However, this process involves a high risk of nuclear proliferation because the recovery of plutonium with a low decontamination factor is difficult in this process. This process involves many high level wastes generates because minor actinide become high level wastes. Toshiba has been developing a new technology, the Toshiba Hybrid Process, for spent-fuel reprocessing based on solvent extraction and pyro-chemical electrolysis, suitable for the
H. Ohmura et al. / Progress in Nuclear Energy 53 (2011) 940e943
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Fig. 3. The method for the advanced extraction.
2. Uranium extraction with electrolytic reduction Fig. 1. Toshiba’s concept of next-generation nuclear fuel cycle.
transition period from LWRs to FRs (Fig. 2) reprocessing because it can produces fuel for FR by reprocessing spent fuel for LWR (Mizuguchi et al., 2008, 2010a). The Hybrid Process combines the solvent extraction process of LWR spent fuel in nitric acid with the recovery of high-purity uranium and the pyro-chemical process in molten salts for the recovery of impure plutonium with minor actinides. High-purity uranium is used for LWR fuel, and impure plutonium with minor actinides is used for metallic FR fuel. The pyro-chemical process for the FR spent fuel recycling system is based on research on an electrorefining process in molten salts conducted at Toshiba since 1988(Koyama et al., 1995; Kobayashi et al., 1997; Koyama et al., 1997; Iizuka et al., 1997; Mizuguchi et al., 2009; Fujita et al., 2008). The new Toshiba Hybrid Process can reduce the amount of high level waste and secondary waste from spent-fuel reprocessing plants, and attain improvement in nuclear proliferation because impure plutonium is recovered with minor actinides for metallic FR fuel. As a first step, an advanced solvent extraction process have been developed in the Toshiba Hybrid Process (Fig. 2)(Mizuguchi et al., 2008, 2010a). The solvent extraction test with electrolytic reduction tests using actual LWR spent fuel were carried out to investigate the uranium recovery purity (target impurity of uranium is less than 1.0 106). As a second step, an Advanced Aqua-Pyro Process have been developed. In this process, actinides and rare earth elements are recovered from an aqueous solution in the extraction process by an oxalic acid precipitation method and are applied to the pyro-chemical electrolysis process. Oxalate acid precipitation tests were carried out in order to evaluate the recovery yield of Pu þ MA (target yield 99.9%).
Fig. 2. Toshiba Hybrid Process based on advanced solvent extraction and pyroelectrolysis.
An advanced solvent extraction process in the Hybrid Reprocessing is shown in Fig. 2. The key technology in this process is the method for the uranium extraction with electrolytic reduction shown in Fig. 3. This technology is based on the PUREX process using 30% tributyl phosphate (TBP) with 70% n-dodecane(n-DD) as an extractant. The method has a single extraction cycle consisting of extraction and electrolytic reduction. This method can extract only uranium by controlling the valence of plutonium ions to the trivalent state by electrolysis. Plutonium is kept in an aqueous solution with minor actinides and can be recovered with minor actinides in the pyro-chemical electrolysis. The hybrid process has high nuclear proliferation resistance because the recovery of plutonium with a high decontamination factor is difficult. Fig. 4 shows the scheme of the uranium extraction with electrolytic reduction. First, uranyl ions (UO2þ 2 ) are reduced to tetravalent uranous (U4þ) ions by electrolysis, and then these uranous ions act as a reductant to reduce tetravalent plutonium ions (Pu4þ) to trivalent ions (Pu3þ). Hexavalent plutonium ions are reduced to trivalent plutonium by tetravalent uranous ions. In the process, the reductant uranous ions are re-oxidized to uranyl ions (Mizuguchi et al., 2008, 2010a,b). Gram-scale actual spent fuel examinations had been carried out, and basic performance of the uranium extraction with electrolytic reduction was confirmed. In view of the gram-scale examinations results, continuous performance tests were carried out, and the impurity of recovered uranium was evaluated. A consecutive processing equipment shown in Fig. 5 was manufactured for the solvent extraction process. Continuous actual spent fuel examinations were carried out using the consecutive processing equipment. The equipment consists of a flow-typed electronic cell, a centrifugal contractor, and four pumps. The cell consists of an anode, a cathode, and a diaphragm made of inorganic material. This cell consists of two compartments.
Fig. 4. The scheme of the uranium extraction with electro-reduction.
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Fig. 6. Oxalate precipitation test using simulated spent fuel.
3. Advanced aqua-pyro process
Fig. 5. Consecutive processing equipment for the solvent extraction process.
The experimental conditions are shown in Table 1. These experiments were carried out at 0.5 mA/cm2. The burnup of actual spent fuel is about 65.9 GWd/tU. It was used as 0.5M-UO2þ 2 in 1M HNO3 solution. The results are given in Table 2. As result, the impurity of recovered uranium was decreased from 1.2 102 to about 1/70 into 1.7 104. The result suggests that the impurity of recovered uranium achieved less than the target value of 1.0 106 in a few stages. It was confirmed that uranium could be recovered with high decontamination factors.
At first an aqua-pyro process was developed (Akai and Fujita, 1995; Akai and Fujita, 1996; Akai and Fujita, 1997; Akai and Fujita, 1998a; Fujita and Akai, 1998b) in which minor actinides (MAs) are recovered from high-level radioactive liquid waste (HLLW) generated by operation of the Rokkasho Reprocessing Plant. This process is an electrolysis technology cultivated from pyro-chemical reprocessing. Next, electrolysis technology based on this aqua-pyro process was applied to an active reprocessing process, to realize a reprocessing process that makes it difficult to separate pure plutonium, while reducing the amount of HLLW. This process introduces reprocessing for LWR fuel that ensures a high level of support for nuclear non-proliferation. Subsequently, pyrochemical reprocessing technology based on electrolysis will be applied to reprocessing of FR fuel. This approach will allow progression of the core technology from the coexistence of FRs and LWRs through to complete adoption of FRs. As a result, Toshiba will be able to offer a fuel cycle suited to both LWRs and FRs in turn. An Advanced Aqua-Pyro Process (Mizuguchi et al., 2008, 2010a,b; Ohmura et al., 2010) has been developed by applying the aqua-pyro process in order to connect the advanced uranium extraction process with the pyroelectrolysis process. In this process, actinides and rare earth elements are recovered from the aqueous solution from the extraction process by the oxalic acid precipitation method and are applied to the pyro-chemical process. Basic oxalate precipitation tests were carried out on simulated spent fuel solution. Fig. 6 shows the test setup. The experimental conditions of the basic tests are shown in Table 3. Precipitation test with neodymium which is simulated as Pu þ MA in nitric acid was carried out because chemical nature of neodymium is similar to TRUs (Akai and Fujita, 1995; Akai and Fujita, 1996; Akai and Fujita, 1997; Akai and Fujita, 1998a; Fujita and Akai, 1998b). Uranium and neodymium were precipitated as uranyl oxalate and neodymium oxalate by the addition of oxalic acid. In the experiments at amount of oxalate ¼ 6 or more, the precipitation ratios of neodymium achieved the target value of 99% (Fig. 7). According to the reference about precipitation tests of rare earth elements by oxalic acid, solubility of neodymium oxalate is comparable as that of other rare earth oxalates. And according to
Table 2 Experimental results.
Table 3 Experimental conditions of basic test.
Table 1 Experimental conditions. 2þ
Concentration of SF
0.5M UO2
Current Flux of SF through electronic cell Flux of SF through bypass line Flux of 30% TBP with 70% n-DD
10A 10 ml/min 90 ml/min 100 ml/min
in 1M-HNO3aq
After supplying actual spent fuel to the cell and bypass line of the cell, a DC electric potential was applied to electrodes to cause 4þ 4þ in the actual spent the reaction of uranium from UO2þ 2 to U . Pu fuel reduce to Pu3þ by U4þ at a point which actual spent fuel through the cell and the actual spent fuel through the bypass line 4þ at the mix. U4þ in the actual spent fuel oxidized to UO2þ 2 by Pu mixing point at the same time. The actual spent fuel after the reaction sent to the centrifugal contractor. And then supplying 30% TBP with 70% n-DD to the centrifugal contractor, much uranium in the actual spent fuel was extracted by the 30% TBP with 70% n-DD, separated into the actual spent fuel and 30% TBP with 70% n-DD. The decontamination factors DFM were calculated according to the following Eq. (1), where C is the concentration of elements.
DFM ¼
CMfeed ,CUTBP CUfeed ,CMTBP
ðM ¼ Pu; Np; Am; CmÞ
(1)
DF Pu MA
682 Np Am Cm
37 953 74
Temperature Amount of oxalate Reaction time Concentration of U Concentration of Nd Concentration of HNO3aq
70 C 1e20 times the stoichiometric 0.5 h 0.1 M 0.033 M 0.1 N
H. Ohmura et al. / Progress in Nuclear Energy 53 (2011) 940e943
943
Table 4 Experimental results of the bench-scale tests. Precipitation ratio Moisture ratio of the precipitation
99.9% or more 4.3%
in argon atmosphere and its moisture is removed before sending to molten salt It is enough to dry to the grade which can be treated as a solid in the advanced aqua-pyro process. Since neodymium could be recovered 99%, it is thought that it is recoverable 99% also about actinides or rare earth elements in actual spent fuel. 4. Conclusion
Fig. 7. Oxalate precipitation test using simulated spent fuel.
the reference, solubility of americium, curium, and plutonium oxalate are about 10 mg/L( ¼ 4 105 M), about 10 mg/L( ¼ 4 105 M), and about 100 mg/L( ¼ 4 104 M) respectively (Rudisill, 1996). Americium, curium, and plutonium contain in actual spent fuel by the same concentration as our experiments, 99% or more of these actinides would be recover in the same conditions. Since neodymium could be recovered 99%, it is thought that it is recoverable 99% also about actinides and rare earth elements in actual spent fuel. It was confirmed to recover 99% of actinides and rare earth elements from simulated spent fuel solution. A bench-scale apparatus shown in Fig. 8 was manufactured for the pyro-chemical process. The bench-scale apparatus consists of a reactor for oxalate precipitation, a solideliquid separator in which nitric acid with fission products and precipitation are separated, and a drying equipment in which the precipitation is dry. It was confirmed that precipitation ratio of neodymium was more than 99.9% and that moisture ratio of the precipitation was less than 10%(Table 4). After this process, the precipitation is further heated
Fig. 8. Bench-scale apparatus for the pyro-chemical process.
Toshiba has proposed a new fuel cycle concept for the transition period from LWRs to FRs. We have been developing a new technology, the Toshiba Hybrid Process, based on solvent extraction and a pyro-chemical process for spent-fuel reprocessing in this transition period. The Toshiba Hybrid Reprocessing Process could recover high-purity U used for LWR fuel, and impure Pu þ MA used for metallic FR fuel. References Akai, Y., Fujita, R., 1995. Development of transuranium element recovery from high level radioactive liquid waste. J. Nucl. Sci. Technol. 32 (10), 807e809. Akai, Y., Fujita, R., 1996. Development of transuranium element recovery from highlevel radioactive liquid waste - conversion of TRU oxalates to chlorides. J. Nucl. Sci. Technol. 33 (10). Akai, Y., Fujita, R., 1997. Development of Transuranium element recovery from highlevel radioactive liquid waste. In: Proc. Int. Conf. on Future Nuclear Systems, Global’97, Yokohama, Japan, Oct. 5-10, 1997. Fujita, R., Akai, Y., 1998a. Development of transuranium elements recovery from high-level radioactive liquid waste. J. Alloys Compounds, 271e273. Fujita, R., Akai, Y., 1998b. Separation of transuranium elements (TRUs) from rare earth elements by electrorefining in molten salts. In: Proc. 49th Annu. Meet. Int. Soc. Electrochemistry, Kokura, Japan, Sept. 15e17. Fujita, R., Mizuguchi, K., Fuse, K., Saso, M., Utsunomiya, K., Abe, K., 2008. Advanced hybrid process with solvent Extraction and pyro-chemical process of spent fuel reprocessing for LWR to FBR. In: 16Th Pac. Basin Nucl. Conf. (16PBNC), Aomori, Japan, Oct. 13-18. Iizuka, M., Koyama, T., Kondo, N., Fujita, R., Tanaka, H., 1997. Actinides recovery from molten salt/liquid metal system by electrochemical methods. J. Nucl. Mater. 247, 183e190. Kobayashi, T., Fujita, R., Nakamura, H., Koyama, T., 1997. Evaluation of cadmium pool potential in an electrorefiner with ceramic partition for spent metallic fuel. J. Nucl. Sci. Technol. 34 (1), 50e57. Koyama, T., Fujita, R., Iizuka, M., Sumida, Y., 1995. Pyrometallurgical reprocessing of fast reactor metallic fuel, -development of a new electrorefiner with a ceramic partition. Nucl. Tech. 110 (6), 357e368. Koyama, T., Iizuka, M., Shoji, Y., Fujita, R., Tanaka, H., Kobayashi, T., Tokiwai, M., 1997. An experimental study of molten salt electrorefining of uranium using solid Iron cathode and liquid cadmium cathode for development of pyrometallurgical reprocessing. J. Nucl. Sci. Technol. 34 (3), 384e393. Mizuguchi K., Fuse K., Nakamura H., Fujita R., Omori T., Utsunomiya K. "Development of Hybrid Reprocessing Technology with Solvent Extraction and Pyrochemical Electrolysis," Proc. Nihon-Genshiryoku-Gakkai, Kochi, Japan, (Sep. 4e6e2008), [in Japanese]. Mizuguchi, K., Fuse, K., Kanamura, S., Fujita, R., Omori, T., Utsunomiya, K., 2009. Development of hybrid reprocessing technology based on solvent extraction and pyrochemical electrolysis. In: Proc. International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities FR09, Kyoto, Japan, Dec. 9. Mizuguchi K., Ohmura H., Kanamura S., Fuse K., Nakamura H., Omori T., Utsunomiya K., Fujita R. "Development of Hybrid Reprocessing Technology (II)," Proc. NihonGenshiryoku-Gakkai, Ibaraki, Japan, (Mar. 26e28e2010a), [in Japanese]. Mizuguchi K., Ohmura H., T. Omori, K. Utsunomiya " Development of Hybrid Reprocessing Technology (I)," Proc. Nihon-Genshiryoku-Gakkai, Ibaraki, Japan, (Mar. 26e28e2010b), [in Japanese]. Ohmura H., Kanamura S., Fuse K., Mizuguchi K., Nakamura H., Omori T., Utsunomiya k., Fujita R. "Development of Hybrid Reprocessing Technology (III)," Proc. Nihon-Genshiryoku-Gakkai, Ibaraki, Japan, (Mar. 26e28e2010), [in Japanese] Rudisill, T.S., 1996. Pretreatment of Americium/Curium Solutions for Vitrification(U). Westinghouse Savannah River Co., Aiken, SC. WSRC-TR-96e0074.