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Journal Articles

Oxygen potential of neodymium-doped U$$_{0.817}$$Pu$$_{0.180}$$Am$$_{0.003}$$O$$_{2 pm x}$$ uranium-plutonium-americium mixed oxides at 1573, 1773, and 1873 K

Vauchy, R.; Sunaoshi, Takeo*; Hirooka, Shun; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Materials, 580, p.154416_1 - 154416_11, 2023/07

Journal Articles

Toward long-term storage of nuclear materials in MOX fuels fabrication facility

Hirooka, Shun; Nakamichi, Shinya; Matsumoto, Taku; Tsuchimochi, Ryota; Murakami, Tatsutoshi

Frontiers in Nuclear Engineering (Internet), 2, p.1119567_1 - 1119567_7, 2023/03

Storage of plutonium (Pu)-containing materials requires extremely strict attention in terms of physical safety and material accounting. Despite the emphasized importance of storage management, only a few reports are available in the public, e.g., experience in PuO$$_{2}$$ storage in the UK and safety standards in the storage of Pu-containing materials in the US. Japan also stores more U-Pu mixed oxide (MOX) mostly in powder form. Adopting an appropriate storage management is necessary depending on the characteristics of MOX items such as raw powder obtained by reprocessing of spent Light Water Reactor fuels, research and development on the remains of fuel fabrication, which can contain organic materials, and dry-recycled powder during fuel fabrication. Stagnation in fuel fabrications and experience in degradation of MOX containers during extended period of storage have led to the review of the storage method in the Plutonium Fuel Development Center in Japan Atomic Energy Agency. The present work discusses the various nuclear materials, storage methods, experience in degradation of containers that occur during storage, and strategies for future long-term storage.

JAEA Reports

Development of "MOX weighing and Ball-mill blending" based on experience in operation and maintenance of MOX fuel manufacturing equipment

Kawasaki, Kohei; Ono, Takanori; Shibanuma, Kimikazu; Goto, Kenta; Aita, Takahiro; Okamoto, Naritoshi; Shinada, Kenta; Ichige, Hidekazu; Takase, Tatsuya; Osaka, Yuki; et al.

JAEA-Technology 2022-031, 91 Pages, 2023/02


The document for back-end policy opened to the public in 2018 by Japan Atomic Energy Agency (hereafter, JAEA) states the decommissioning of facilities of Nuclear Fuel Cycle Engineering Laboratories and JAEA have started gathering up nuclear fuel material of the facilities into Plutonium Fuel Production Facilities (hereafter, PFPF) in order to put it long-term, stable and safe storage. Because we planned to manufacture scrap assemblies almost same with Monju fuel assembly using unsealed plutonium-uranium mixed-oxide (hereafter, MOX) powder held in PFPF and transfer them to storage facilities as part of this "concentration" task of nuclear fuel material, we obtained permission to change the use of nuclear fuel material in response to the new regulatory Requirements in Japan for that. The amount of plutonium (which is neither sintered pellets nor in a lidded powder-transport container) that could be handled in the pellet-manufacturing process was limited to 50 kg Pu or less in order to decrease the facility risk in this manufacture. Therefore, we developed and installed the "MOX weighing and blending equipment" corresponding with small batch sizes that functioned in a starting process and the equipment would decrease handling amounts of plutonium on its downstream processes. The failure data based on our operation and maintenance experiences of MOX fuel production facilities was reflected in the design of the equipment to further improve reliability and maintainability in this development. The completed equipment started its operation using MOX powder in February 2022 and the design has been validated through this half-a-year operation. This report organizes the knowledge obtained through the development of the equipment, the evaluation of the design based on the half-a-year operation results and the issues in future equipment development.

Journal Articles

Oxygen diffusion in the fluorite-type oxides CeO$$_{2}$$, ThO$$_{2}$$, UO$$_{2}$$, PuO$$_{2}$$, and (U, Pu)O$$_{2}$$

Kato, Masato; Watanabe, Masashi; Hirooka, Shun; Vauchy, R.

Frontiers in Nuclear Engineering (Internet), 1, p.1081473_1 - 1081473_10, 2023/01

Journal Articles

Computer code analysis of irradiation performance of an annular mixed oxide fuel element

Yokoyama, Keisuke; Uwaba, Tomoyuki

Journal of Nuclear Science and Technology, 9 Pages, 2023/00

 Times Cited Count:0

no abstracts in English

Journal Articles

Cation interdiffusion in uranium-plutonium mixed oxide fuels; Where are we now?

Vauchy, R.; Hirooka, Shun; Matsumoto, Taku; Kato, Masato

Frontiers in Nuclear Engineering (Internet), 1, p.1060218_1 - 1060218_18, 2022/12

Journal Articles

Materials science and fuel technologies of uranium and plutonium mixed oxide

Kato, Masato; Machida, Masahiko; Hirooka, Shun; Nakamichi, Shinya; Ikusawa, Yoshihisa; Nakamura, Hiroki; Kobayashi, Keita; Ozawa, Takayuki; Maeda, Koji; Sasaki, Shinji; et al.

Materials Science and Fuel Technologies of Uranium and Plutonium mixed Oxide, 171 Pages, 2022/10

Innovative and advanced nuclear reactors using plutonium fuel has been developed in each country. In order to develop a new nuclear fuel, irradiation tests are indispensable, and it is necessary to demonstrate the performance and safety of nuclear fuels. If we can develop a technology that accurately simulates irradiation behavior as a technology that complements the irradiation test, the cost, time, and labor involved in nuclear fuel research and development will be greatly reduced. And safety and reliability can be significantly improved through simulation of nuclear fuel irradiation behavior. In order to evaluate the performance of nuclear fuel, it is necessary to know the physical and chemical properties of the fuel at high temperatures. And it is indispensable to develop a behavior model that describes various phenomena that occur during irradiation. In previous research and development, empirical methods with fitting parameters have been used in many parts of model development. However, empirical techniques can give very different results in areas where there is no data. Therefore, the purpose of this study is to construct a scientific descriptive model that can extrapolate the basic characteristics of fuel to the composition and temperature, and to develop an irradiation behavior analysis code to which the model is applied.

Journal Articles

Measurements of thermal conductivity for near stoichiometric (U$$_{0.7-z}$$Pu$$_{0.3}$$Am$$_{z}$$)O$$_{2}$$ (z = 0.05, 0.10, and 0.15)

Yokoyama, Keisuke; Watanabe, Masashi; Tokoro, Daishiro*; Sugimoto, Masatoshi*; Morimoto, Kyoichi; Kato, Masato; Hino, Tetsushi*

Nuclear Materials and Energy (Internet), 31, p.101156_1 - 101156_7, 2022/06

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

In current nuclear fuel cycle systems, to reduce the amount of high-level radioactive waste, minor actinides (MAs) bearing MOX fuel is one option for burning MAs using fast reactor. However, the effects of Am content in fuel on thermal conductivity are unclear because there are no experimental data on thermal conductivity of high Am bearing MOX fuel. In this study, The thermal conductivities of near stoichiometric (U$$_{0.7-z}$$Pu$$_{0.3}$$Am$$_{z}$$)O$$_{2}$$ solid solutions(z = 0.05, 0.10, and 0.15) have been measured between room temperature (RT) and 1473 K. The thermal conductivities decreased with increasing Am content and satisfied the classical phonon transport model ((A+BT)$$^{-1}$$) up to about 1473 K. A values increased linearly with increasing Am content because the change in ionic radius affects the conduction of the phonon due to the solid solution in U$$^{5+}$$ and Am$$^{3+}$$. B values were independent of Am content.

Journal Articles

Recent studies on fuel properties and irradiation behaviors of Am/Np-bearing MOX

Hirooka, Shun; Yokoyama, Keisuke; Kato, Masato

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04

Property studies on Am/Np-bearing MOX were carried out and how the properties influences on the irradiation behaviors was discussed. Both Am and Np inclusions increase the oxygen potential of MOX. Inter-diffusion coefficients obtained by using diffusion couple technique indicate that the inter-diffusion coefficient is larger in the order of U-Am, U-Pu and U-Np. Also, the inter-diffusion coefficients were evaluated to be larger at the O/M = 2 than those of O/M $$<$$ 2 by several orders. The increase of oxygen potential with Am/Np leads to higher vapor pressure of UO$$_{3}$$ and the acceleration of the pore migration along temperature gradient during irradiation. The redistributions of actinide elements were also considered with the relationship of the pore migration and diffusion in solid state. Thus, the obtained inter-diffusion coefficients directly influence on the redistribution rate. The obtained properties were modelled and can be installed in a fuel irradiation simulation code.

Journal Articles

Advanced reactor experiments for sodium fast reactor fuels (ARES) project; Transient irradiation experiments for metallic and MOX fuels

Jensen, C. B.*; Wachs, D. M.*; Woolstenhulme, N. E.*; Ozawa, Takayuki; Hirooka, Shun; Kato, Masato

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

Journal Articles

Development of fission gas release model for MOX fuel pellets with treatment of heterogeneous microstructure

Tasaki, Yudai; Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 59(3), p.382 - 394, 2022/03

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

Journal Articles

Relative oxygen potential measurements of (U,Pu)O$$_{2}$$ with Pu = 0.45 and 0.68 and related defect formation energy

Hirooka, Shun; Matsumoto, Taku; Sunaoshi, Takeo*; Hino, Tetsushi*

Journal of Nuclear Materials, 558, p.153375_1 - 153375_8, 2022/01

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Impact of MOX fuel use in light-water reactors; Long-term radiological consequences of disposal of high-level waste in a geological repository

Minari, Eriko*; Kabasawa, Satsuki; Mihara, Morihiro; Makino, Hitoshi; Asano, Hidekazu*; Nakase, Masahiko*; Takeshita, Kenji*

Journal of Nuclear Science and Technology, 11 Pages, 2022/00

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

Journal Articles

Data processing and visualization of X-ray computed tomography images of a JOYO MK-III fuel assembly

Tsai, T.-H.; Sasaki, Shinji; Maeda, Koji

Journal of Nuclear Science and Technology, 9 Pages, 2022/00

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

Journal Articles

Development of fuel performance analysis code, BISON for MOX, named Okami; Analyses of pore migration behavior to affect the MA-bearing MOX fuel restructuring

Ozawa, Takayuki; Hirooka, Shun; Kato, Masato; Novascone, S.*; Medvedev, P.*

Journal of Nuclear Materials, 553, p.153038_1 - 153038_16, 2021/09

 Times Cited Count:2 Percentile:48.75(Materials Science, Multidisciplinary)

To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with INL and have installed pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation experiment in Joyo was evaluated by the 2-dimensional analyses. Four MA-bearing MOX pins with different O/M ratio and pellet/cladding gap size were irradiated in Joyo B14 experiment. Remarkable restructuring of stoichiometric MA-bearing MOX fuels was observed in PIE, and could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void assumes to move toward wide-gap side when the pellet eccentricity taking place, but 2-dimentional analyses on pellet transverse section revealed that the central void formation observed in PIE would be inconsistent with a direction of the pellet eccentricity.

JAEA Reports

Improvement of intragranular fission gas behavior model for fuel performance code FEMAXI-8

Udagawa, Yutaka; Tasaki, Yudai

JAEA-Data/Code 2021-007, 56 Pages, 2021/07


Japan Atomic Energy Agency (JAEA) has released FEMAXI-8 in 2019 as the latest version of the fuel performance code FEMAXI, which has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly normal operation conditions and anticipated transient conditions. This report summarizes a newly developed model to analyze intragranular fission gas behaviors considering size distribution of gas bubbles and their dynamics. While the intragranular fission gas behavior models implemented in the previous FEMAXI versions have supported only treating single bubble size for a given fuel element, the new model supports multiple gas groups according to their size and treats their dynamic behaviors, making the code more versatile. The model was first implemented as a general module that takes arbitrary number of bubble groups and spatial mesh division for a given fuel grain system. An interface module to connect the model to FEMAXI-8 was then developed so that it works as a 2 bubble groups model, which is the minimum configuration of the multi-grouped model to be operated with FEMAXI-8 at the minimum calculation cost. FEMAXI-8 with the new intragranular model was subjected to a systematic validation calculation against 144 irradiation test cases and recommended values for model parameters were determined so that the code makes reasonable predictions in terms of fuel center temperature, fission gas release, etc. under steady-state and power ramp conditions.

Journal Articles

Preventing nuclear fuel material adhesion on glove box components using nanoparticle coating

Segawa, Tomoomi; Kawaguchi, Koichi; Ishii, Katsunori; Suzuki, Masahiro; Tachihara, Joji; Takato, Kiyoto; Okita, Takatoshi; Satone, Hiroshi*; Suzuki, Michitaka*

Mechanical Engineering Journal (Internet), 8(3), p.21-00022_1 - 21-00022_9, 2021/06

To reduce the hold-up of the nuclear fuel materials in the glove box and the external exposure dose, the technology of the MOX powder adhesion prevention by the nanoparticle coating to the acrylic panels of the glove box has been developed. The surface analysis by means of atomic force microscopy (AFM) showed that the acrylic test piece surface coated with nanoparticles had a higher root mean square roughness value than that non-coated with nanoparticles. Due to the formation of nano-sized tiny rugged surface, the nanoparticle coating reduced the minimum adhesion force between the UO$$_{2}$$ particles and the acrylic test piece surface with the smallest particle size of about 5 $$mu$$m where desorption was observed, by about one-tenth. Moreover, the nanoparticle coating reduced the amount of the MOX powder adhering to the acrylic test piece to about one-tenth. In this study, it was found that applying the nanoparticle coating to the acrylic panels of glove box can prevent the adhesion of nuclear fuel materials. This method is effective for reducing the hold-up of the nuclear fuel materials in the glove box, the external exposure dose and improving the visibility of the acrylic panels.

Journal Articles

Leaching behavior of radionuclides from samples prepared from spent fuel rod comparable to core debris in the 1F NPS

Onishi, Takashi; Maeda, Koji; Katsuyama, Kozo

Journal of Nuclear Science and Technology, 58(4), p.383 - 398, 2021/04

 Times Cited Count:8 Percentile:81.16(Nuclear Science & Technology)

Journal Articles

Present status and practical issues on dosimetry for the lens of the eye at JAEA MOX Fuel Facilities

Tsujimura, Norio; Yamazaki, Takumi; Takada, Chie

Journal of Nuclear Science and Technology, 58(1), p.40 - 44, 2021/01

 Times Cited Count:3 Percentile:45.37(Nuclear Science & Technology)

Journal Articles

Oxygen potential measurement of (U,Pu,Am)O$$_{2 pm x}$$ and (U,Pu,Am,Np)O$$_{2 pm x}$$

Hirooka, Shun; Matsumoto, Taku; Kato, Masato; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*

Journal of Nuclear Materials, 542, p.152424_1 - 152424_9, 2020/12

 Times Cited Count:3 Percentile:45.37(Materials Science, Multidisciplinary)

The measurement of oxygen potential was conducted at 1,673, 1,773, and 1,873 K for (U$$_{0.623}$$Pu$$_{0.350}$$Am$$_{0.027}$$)O$$_{2}$$ and at 1,873 and 1,923 K for (U$$_{0.553}$$Pu$$_{0.285}$$Am$$_{0.015}$$Np$$_{0.147}$$)O$$_{2}$$ by using a thermo-gravimeter and an oxygen sensor. Am inclusion in terms of substituting the U significantly increased the oxygen potential. Similarly, the inclusion of Np as a substitute for U increased the oxygen potential; however, the effect was not as large as that with the Pu or Am addition at the same rate. The results were analyzed via defect chemistry and certain defect formations were suggested in the reducing region and the near-stoichiometric region by plotting the relationship between PO$$_{2}$$ and the deviation from the stoichiometry. The equilibrium constants of the defect reactions were arranged to reproduce the experiment such that Am/Np contents were included in the entropy with coefficients fitting the experimental data.

298 (Records 1-20 displayed on this page)