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Journal Articles

Effect of decay heat on pyrochemical reprocessing of minor actinide transmutation nitride fuels

Hayashi, Hirokazu; Tsubata, Yasuhiro; Sato, Takumi

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(3), p.97 - 107, 2023/08

The Japan Atomic Energy Agency has chosen nitride fuel as the first candidate for the transmutation of long-lived minor actinides (MA) using accelerator-driven systems (ADS). The pyrochemical method has been considered for reprocessing spent MA nitride fuels, because their decay heat should be very large for aqueous reprocessing. This study was conducted to investigate the effect of decay heat on the pyrochemical reprocessing of MA nitride fuels. On the basis of the estimated decay heats and the temperature limits of the materials that are to be handled in pyrochemical reprocessing, quantities adequate for handling in argon gas atmosphere were evaluated. From these considerations, we proposed that an electrorefiner with a diameter of 26 cm comprising 12 cadmium (Cd) cathodes with a diameter of 4 cm is suitable. On the basis of the size of the electrorefiner, the number necessary to reprocess spent MA fuels from 1 ADS in 200 days was evaluated to be 25. Furthermore, the amount of Cd-actinides (An) alloy to produce An nitrides by the nitridation-distillation combined reaction process was proposed to be about one-quarter that of Cd-An cathode material. The evaluated sizes and required numbers of equipment support the feasibility of pyrochemical reprocessing for MA nitride fuels.

Journal Articles

Kinetic mass transfer behavior of Eu(III) in nitrilotriacetamide-impregnated polymer-coated silica particles

Miyagawa, Akihisa*; Hayashi, Naoki*; Kuzure, Yoshiaki*; Takahashi, Takumi*; Iwamoto, Hibiki*; Arai, Tsuyoshi*; Nagatomo, Shigenori*; Miyazaki, Yasunori; Hasegawa, Kenta; Sano, Yuichi; et al.

Bulletin of the Chemical Society of Japan, 96(7), p.671 - 676, 2023/07

We investigated the distribution mechanism of Eu(III) in a single polymer-coated silica particle including nitrilotriacetamide (NTA) extractants known as HONTA and TOD2EHNTA. The present study provides a valuable approach for the evaluation and enhancement of the functionality of "single extractant-impregnated polymer-coated silica particle".

Journal Articles

Formation of MPd$$_{3+x}$$ (M = Gd, Np) by the reaction of MN with Pd and chlorination of MPd$$_{3+x}$$ using cadmium chloride

Hayashi, Hirokazu; Shibata, Hiroki; Sato, Takumi; Otobe, Haruyoshi

Journal of Radioanalytical and Nuclear Chemistry, 332(2), p.503 - 510, 2023/02

 Times Cited Count:0 Percentile:0.01(Chemistry, Analytical)

The formation of MPd$$_{3+x}$$ (M = Gd, Np) by the reaction of MN with Pd at 1323 K in Ar gas flow was observed. Cubic AuCu$$_3$$-type GdPd$$_{3.3}$$ (${it a}$ = 0.4081 $$pm$$ 0.0001 nm) and NpPd$$_3$$ (${it a}$ = 0.4081 $$pm$$ 0.0001 nm) were identified, respectively. The product obtained from the reaction of NpN with Pd contained additional phases including the hexagonal TiNi$$_3$$-type NpPd$$_3$$. Chlorination of the MPd$$_{3+x}$$ (M = Gd, Np) samples was accomplished by the solid-state reaction using cadmium chloride at 673 K in a dynamic vacuum. Pd-rich solid solution phase saturated with Cd and an intermetallic compound PdCd were obtained as by-products of MCl$$_3$$ formation.

Journal Articles

Material balance evaluation of pyroprocessing for minor actinide transmutation nitride fuel

Tateno, Haruka; Sato, Takumi; Tsubata, Yasuhiro; Hayashi, Hirokazu

Journal of Nuclear Science and Technology, 57(3), p.224 - 235, 2020/03

 Times Cited Count:6 Percentile:56.35(Nuclear Science & Technology)

Fuel cycle technology for the transmutation of long-lived minor actinides (MAs) using an accelerator-driven system has been developed using the double-strata fuel cycle concept. A mononitride solid solution of MAs and Pu diluted with ZrN is a prime fuel candidate for the accelerator-driven transmutation of MAs. Pyro-reprocessing is suitable for recycling the residual MAs in irradiated nitride fuel with high radiation doses and decay heat. Spent nitride fuel is anodically dissolved, and the actinides are recovered simultaneously into a liquid cadmium cathode via molten salt electrorefining. The process should be designed to achieve the target recovery yield of MAs and the acceptable impurity level of rare earths in the recovered material. We evaluated the material balance during the pyro-reprocessing of spent nitride fuel to gain important insight on the design process. We examined the effects of changing processing conditions on material flow and quantity of waste.

Journal Articles

Research and development on pyrochemical treatment of spent nitride fuels for MA transmutation in JAEA

Hayashi, Hirokazu; Sato, Takumi; Shibata, Hiroki; Tsubata, Yasuhiro

NEA/NSC/R(2017)3, p.427 - 432, 2017/11

Transmutation of long-lived radioactive nuclides including minor actinides (MA: Np, Am, Cm) has been studied in Japan Atomic Energy Agency (JAEA). Pb-Bi cooled sub-critical accelerator-driven system (ADS) is regarded as one of the powerful tools for transmutation of MA under the double strata fuel cycle concept. Uranium-free MA-Pu nitride fuel was chosen as the first candidate for MA transmutation. Reprocessing of spent ADS fuel and reusing MA recovered from the spent ADS fuels is necessary to improve the transmutation ratio. A pyrochemical process has been proposed as the first candidate for reprocessing of the spent nitride fuel for MA transmutation, because this technique has some advantages over aqueous process, such as the resistance to radiation damage, which is an important issue for the fuels containing large amounts of highly radioactive MA, and feasibility for recovering expensive N-15 in the spent fuels to be reused. This paper overviews the current status of the technology development, including our recent study. Development of the anode suitable for electro-refining of nitride fuels and that of the apparatus for renitridation of the metals recovered in Cd cathode for 100g-Cd scale cold tests are main topics. Evaluation of the batch sizes of each process, which is necessary for estimating the scale of the engineering-apparatus, with considering the decay heat of MA and FP, will also be introduced.

Journal Articles

Effect of hydrocarbons on the efficiency of catalytic reactor of detritiation system in an event of fire

Edao, Yuki; Sato, Katsumi; Iwai, Yasunori; Hayashi, Takumi

Journal of Nuclear Science and Technology, 53(11), p.1831 - 1838, 2016/11

 Times Cited Count:7 Percentile:60.71(Nuclear Science & Technology)

Journal Articles

Deuterium permeation behavior for damaged tungsten by ion implantation

Oya, Yasuhisa*; Li, X.*; Sato, Misaki*; Yuyama, Kenta*; Oyaizu, Makoto; Hayashi, Takumi; Yamanishi, Toshihiko; Okuno, Kenji*

Journal of Nuclear Science and Technology, 53(3), p.402 - 405, 2016/03

 Times Cited Count:10 Percentile:68.73(Nuclear Science & Technology)

The deuterium (D) permeation behaviors for ion damaged tungsten (W) by 3 keV D$$_{2}$$$$^{+}$$ and 10 keV C$$^{+}$$ were studied. The D permeability was obtained for un-damaged W at various temperatures. For both D$$_{2}$$$$^{+}$$ and C$$^{+}$$ implanted W, the permeability was clearly reduced. But, for the D$$_{2}$$$$^{+}$$ implanted W, the permeability was recovered by heating at 1173 K and it was almost consistent with that for un-damaged W. In the case of C$$^{+}$$ implanted W, the permeability was not recovered even if the sample was heated at 1173 K, indicating that the existence of carbon would prevent the recovery of permeation path in W. In addition, TEM observation showed the voids were grown by heating at 1173 K and not removed, showing the existence of damages would not largely influence on the hydrogen permeation behavior in W in the present study.

Journal Articles

R&D activities of tritium technologies on Broader Approach in Phase 2-2

Isobe, Kanetsugu; Kawamura, Yoshinori; Iwai, Yasunori; Oyaizu, Makoto; Nakamura, Hirofumi; Suzuki, Takumi; Yamada, Masayuki; Edao, Yuki; Kurata, Rie; Hayashi, Takumi; et al.

Fusion Engineering and Design, 98-99, p.1792 - 1795, 2015/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Activities on Broader Approach (BA) were started in 2007 on the basis of the Agreement between the Government of Japan and the EURATOM. The period of BA activities consist of Phase1 and Phase2 dividing into Phase 2-1 (2010-2011), Phase 2-2 (2012-2013) and Phase 2-3 (2014-2016). Tritium technology was chosen as one of important R&D issues to develop DEMO plant. R&D activities of tritium technology on BA consist of four tasks. Task-1 is to prepare and maintain the tritium handling facility in Rokkasho BA site in Japan. Task 2, 3 and 4 are main R&D activities for tritium and these are focused on: Task-2) Development of tritium accountancy technology, Task-3) Development of basic tritium safety research, Task-4) Tritium durability test. R&D activities of tritium technology in Phase 2-2 were underway successfully and closed in 2013.

Journal Articles

Chlorination of UO$$_{2}$$ and (U,Zr)O$$_{2}$$ solid solution using MoCl$$_{5}$$

Sato, Takumi; Shibata, Hiroki; Hayashi, Hirokazu; Takano, Masahide; Kurata, Masaki

Journal of Nuclear Science and Technology, 52(10), p.1253 - 1258, 2015/10

 Times Cited Count:6 Percentile:46.16(Nuclear Science & Technology)

In order to explore the applicability of the chlorination by MoCl$$_{5}$$ as a potential pretreatment technique for waste treatment of fuel debris by pyrochemical methods, chlorination experiments of UO$$_{2}$$ and (U$$_{0.5}$$Zr$$_{0.5}$$)O$$_{2}$$ simulated fuel debris were carried out in two steps: the first one is a chlorination reaction by homogeneous heating, the second one is a volatilization of molybdenum by-product by heating under temperature gradient condition. Most of UO$$_{2}$$ and (U$$_{0.5}$$Zr$$_{0.5}$$)O$$_{2}$$ powder were converted to UCl$$_{4}$$ or UCl$$_{4}$$ and ZrCl$$_{4}$$ mixture at 573 K, respectively. In the case of (U$$_{0.5}$$Zr$$_{0.5}$$)O$$_{2}$$sintered particle, most of sample was converted to the chlorides because the products evaporated and be separated from sample surface at 773 K, while only the surface of the sample disk was converted to the chlorides at 573 and 673 K. Most of molybdenum by-product and ZrCl$$_{4}$$ were separated from UCl$$_{4}$$ by volatilization at 573 K.

Journal Articles

Development of nitride fuel cycle technology for transmutation of minor actinides

Hayashi, Hirokazu; Nishi, Tsuyoshi*; Sato, Takumi; Kurata, Masaki

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1811 - 1817, 2015/09

Transmutation of long-lived radioactive nuclides including minor actinides (MA: Np, Am, Cm) has been studied in Japan Atomic Energy Agency (JAEA). Accelerator-driven system (ADS) is regarded as one of the powerful tools for transmutation of MA under the double strata fuel cycle concept. Uranium-free nitride fuel was chosen as the first candidate fuel for MA transmutation using ADS. To improve the transmutation ratio of MA, reprocessing of spent fuel and reusing MA recovered from the spent fuels is necessary. Our target is to transmute 99% of MA arisen from commercial power reactor fuel cycle, with which the period until the radiotoxicity drops below that of natural uranium can be shorten from about 5000 years to about 300 years. A pyrochemical process has been proposed as the first candidate for reprocessing of the spent nitride fuel. This paper overviews the current status of the nitride fuel cycle technology. Our recent study on fuel fabrication, fuel property measurements, reprocessing of spent fuel, development of the property database of MA nitride fuel, and fuel behavior simulation code are introduced. Our research and development (R&D) plan based on the roadmap of the development is also introduced.

Journal Articles

Tracing temperature in a nanometer size region in a picosecond time period

Nakajima, Kaoru*; Kitayama, Takumi*; Hayashi, Hiroaki*; Matsuda, Makoto; Sataka, Masao*; Tsujimoto, Masahiko*; Toulemonde, M.*; Bouffard, S.*; Kimura, Kenji*

Scientific Reports (Internet), 5, p.13363_1 - 13363_8, 2015/08

 Times Cited Count:5 Percentile:41.19(Multidisciplinary Sciences)

Journal Articles

Recent progress and future R&D plan of nitride fuel cycle technology for transmutation of minor actinides

Hayashi, Hirokazu; Nishi, Tsuyoshi; Takano, Masahide; Sato, Takumi; Shibata, Hiroki; Kurata, Masaki

NEA/NSC/R(2015)2 (Internet), p.360 - 367, 2015/06

Uranium-free nitride fuel was chosen as the first candidate for transmutation of long-lived minor actinides (MA) using accelerator-driven system (ADS) under the double strata fuel cycle concept by Japan Atomic Energy Agency (JAEA). The advantages of nitride fuel are good thermal properties and large mutual solubility among actinide elements. A pyrochemical process is proposed as the first candidate for the reprocessing of the spent nitride fuel, because this technique has some advantages over aqueous process, such as the resistance to radiation damage, which is an important issue for the fuels containing large amounts of highly radioactive MA. This paper overviews the recent progress and future R&D plan of the study on the nitride fuel cycle technology in JAEA.

Journal Articles

Effect of tritium on corrosion behavior of chromium in 0.01N sulfuric acid solution

Oyaizu, Makoto; Isobe, Kanetsugu; Hayashi, Takumi

Fusion Science and Technology, 67(3), p.519 - 522, 2015/04

 Times Cited Count:2 Percentile:17.75(Nuclear Science & Technology)

The effects of tritiated water on the corrosion behavior of chromium were electrochemically studied by anodic polarization measurements with changing tritium concentration and dissolved oxygen concentration as parameters in the electrolyte of 0.01N sulfuric acid solution, self-passivation due to dissolved oxygen could be observed in pure water without tritium. As a result, it was found that the self-passivation was inhibited in tritiated electrolyte as shown in the previous studies for SUS304 stainless steel. It is indicated from the result that the passivation inhibitory effect for SUS304 stainless steel could be induced by dissolution of chromium in passivation film on SUS304 stainless steel.

Journal Articles

Recent progress on tritium technology research and development for a fusion reactor in Japan Atomic Energy Agency

Hayashi, Takumi; Nakamura, Hirofumi; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; Yamada, Masayuki; Suzuki, Takumi; Kurata, Rie; Oyaizu, Makoto; Edao, Yuki; et al.

Fusion Science and Technology, 67(2), p.365 - 370, 2015/03

 Times Cited Count:1 Percentile:9.79(Nuclear Science & Technology)

Journal Articles

Measurement of tritium penetration through concrete material covered by various paints coating

Edao, Yuki; Kawamura, Yoshinori; Kurata, Rie; Fukada, Satoshi*; Takeishi, Toshiharu*; Hayashi, Takumi; Yamanishi, Toshihiko

Fusion Science and Technology, 67(2), p.320 - 323, 2015/03

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The present study aims at obtaining fundamental knowledge for tritium transfer behavior and interaction between tritium and paint coated on concrete walls. The amounts of tritium penetration and release in cement paste with epoxy and urethane paint coatings were measured. The tritium penetration amounts were increased with the HTO exposure time. Time to achieve each saturate tritium value was more than 60 days for cement paste coated with epoxy paint and with urethane paint, while cement paste without paint took 2 days to achieve it. Tritium penetration rates were estimated by an analysis of diffusion model. Although their paint coatings were effective for reduction of tritium penetration through the cement paste exposed to HTO for a short period, the amount of tritium trapped in the paints became large for a long time. This work has been performed under the collaboration research between JAEA and Kyushu University.

Journal Articles

Correlation of rates of tritium migration through porous concrete

Fukada, Satoshi*; Katayama, Kazunari*; Takeishi, Toshiharu*; Edao, Yuki; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko

Fusion Science and Technology, 67(2), p.99 - 102, 2015/03

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Pyrochemical treatment of spent nitride fuels for MA transmutation

Hayashi, Hirokazu; Sato, Takumi; Shibata, Hiroki; Kurata, Masaki; Iwai, Takashi; Arai, Yasuo

Science China; Chemistry, 57(11), p.1427 - 1431, 2014/11

 Times Cited Count:5 Percentile:5.9(Chemistry, Multidisciplinary)

Nitride fuels have several advantages, such as high thermal conductivity and high metal density like metallic fuels, and high melting point and isotropic crystal structure like oxide fuels. Since the late 1990s, the partitioning and transmutation of minor actinides (MA) has been studied to decrease the long term radio-toxicity of high level waste and mitigate the burden on the final disposal. Japan Atomic Energy Agency (JAEA) has been proposing dedicated transmutation cycle using the Accelerator-Driven System (ADS) with the nitride fuels containing MA. We have been developing the nitride fuel cycle including pyrochemical process. Our focus is on electrolysis of nitride fuels and refabrication of nitride fuel from the recovered actinides because other processes are similar to the technology for the metal fuel treatment and have been studied elsewhere. In this paper, we summarized our activity on developments of the pyrochemical treatment of the spent nitride fuels.

Journal Articles

R&D status on water cooled ceramic breeder blanket technology

Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.

Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10

 Times Cited Count:20 Percentile:84.35(Nuclear Science & Technology)

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.

Journal Articles

Study of safety features and accident scenarios in a fusion DEMO reactor

Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.

Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10

 Times Cited Count:13 Percentile:70.4(Nuclear Science & Technology)

After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.

Journal Articles

Hydrogen and water vapor adsorption properties on cation-exchanged mordenite for use to a tritium recovery system

Kawamura, Yoshinori; Edao, Yuki; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko

Fusion Engineering and Design, 89(7-8), p.1539 - 1543, 2014/10

 Times Cited Count:7 Percentile:53.55(Nuclear Science & Technology)

Tritium recovery system using adsorption or catalytic isotope exchange has already been proposed for a solid breeding blanket system of a nuclear fusion reactor. Synthetic zeolite is often used as an adsorbent or a substrate of chemical exchange catalyst. And, it is well known that its properties are changed easily by exchanging their cations. So, in this work, adsorption capacities of hydrogen isotope and water vapor on cation-exchanged mordenite with transition metal ion were investigated. Ag ion-exchanged mordenite (Ag-MOR) has indicated considerably large hydrogen adsorption capacity in lower pressure range at 77 K. And, adsorption capacity of water vapor did not so vary with exchaned cation in comparison with hydrogen adsorption. The discussion from the viewpoint of adsorption rate is still remaining, but more compact cryosorption column for tritium recovery system is possible to design if Ag-MOR is adopted.

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