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Journal Articles

Comparison of correlations for thermal creep of FBR MOX

Calabrese, R.*; Hirooka, Shun

Progress in Nuclear Energy, 178, p.105516_1 - 105516_11, 2025/01

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Thermal creep is one of the key properties of mixed oxide (MOX) fuel for innovative fast reactors. Thermal creep of fuel affects markedly the interaction between the fuel and the cladding. A review of correlations available in the literature is presented. The effect of porosity, plutonium concentration, and stoichiometry are discussed also in the light of recent numerical results. Our analysis pointed out some inconsistencies concerning the modelling of the effect of porosity on diffusional creep and a re-evaluation of the effect of plutonium concentration. The discussion suggested that Evans's findings on the effect of stoichiometry should be better assessed as well as the level of increase in creep moving towards stoichiometry. Typical operating conditions of fast breeder reactors confirmed the need for an extension of porosity and temperature correlations' domains. Besides this, a new correlation based on a separate-effect approach has been proposed for fuel performance codes.

Journal Articles

Uranium-plutonium-oxygen phase diagram; Investigating the solvus of fluorite's exsolution

Vauchy, R.; Hirooka, Shun; Horii, Yuta; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Yamada, Tadahisa*; Tamura, Tetsuya*; Murakami, Tatsutoshi

Journal of Nuclear Materials, 599, p.155233_1 - 155233_11, 2024/10

 Times Cited Count:0 Percentile:0.00(Materials Science, Multidisciplinary)

The fluorite exsolution/recombination in U$$_{1-y}$$Pu$$_{y}$$O$$_{2-x}$$ (y = 0.30 and 0.45) and PuO$$_{2-x}$$ was investigated using differential scanning calorimetry. The results are in relatively good agreement with the literature data, except for plutonia. Our values indicate that the critical temperature of the miscibility gap in Pu-O is 30$$sim$$50 K lower than previously reported. Finally, the systematic experimental procedure allowed us refining the locus of the solvus existing in hypostoichiometric U$$_{0.70}$$0Pu$$_{0.30}$$O$$_{2-x}$$, U$$_{0.55}$$Pu$$_{0.45}$$O$$_{2-x}$$, and PuO$$_{2-x}$$ dioxides.

Journal Articles

Sintering behavior analysis of compacted dry recycled U$$_{0.7}$$Pu$$_{0.3}$$O$$_{2}$$ powder using master sintering curve theory

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

Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07

 Times Cited Count:1 Percentile:51.66(Materials Science, Multidisciplinary)

Journal Articles

Numerical study of initiating phase of core disruptive accident in small sodium-cooled fast reactors with negative void reactivity

Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi

Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05

 Times Cited Count:1 Percentile:23.64(Nuclear Science & Technology)

Journal Articles

A Science-based mixed oxide property model for developing advanced oxide nuclear fuels

Kato, Masato; Oki, Takumi; Watanabe, Masashi; Hirooka, Shun; Vauchy, R.; Ozawa, Takayuki; Uwaba, Tomoyuki; Ikusawa, Yoshihisa; Nakamura, Hiroki; Machida, Masahiko

Journal of the American Ceramic Society, 107(5), p.2998 - 3011, 2024/05

 Times Cited Count:5 Percentile:41.38(Materials Science, Ceramics)

Journal Articles

High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka

Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01

 Times Cited Count:1 Percentile:0.00(Nuclear Science & Technology)

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

 Times Cited Count:9 Percentile:93.25(Materials Science, Multidisciplinary)

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, 60(6), p.715 - 723, 2023/06

 Times Cited Count:1 Percentile:14.04(Nuclear Science & Technology)

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

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:6 Percentile:67.31(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

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

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

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:0.00(Nuclear Science & Technology)

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

JAEA-Data-Code-2021-007.pdf:5.05MB

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

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:10 Percentile:68.22(Nuclear Science & Technology)

Journal Articles

Post-irradiation examinations of annular mixed oxide fuels with average burnup 4 and 5% FIMA

Cappia, F.*; Tanaka, Kosuke; Kato, Masato; McClellan, K.*; Harp, J.*

Journal of Nuclear Materials, 533, p.152076_1 - 152076_14, 2020/05

 Times Cited Count:10 Percentile:68.22(Materials Science, Multidisciplinary)

Journal Articles

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

 Times Cited Count:10 Percentile:58.28(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

Taniguchi, Yoshinori; Udagawa, Yutaka; Mihara, Takeshi; Amaya, Masaki; Kakiuchi, Kazuo

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

Journal Articles

Model updates and performance evaluations on fuel performance code FEMAXI-8 for light water reactor fuel analysis

Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06

 Times Cited Count:12 Percentile:69.57(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

73 (Records 1-20 displayed on this page)