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Yokoyama, Keisuke; Watanabe, Masashi; Onishi, Takashi; Yano, Yasuhide; Tokoro, Daishiro*; Sugata, Hiromasa*; Kato, Masato*
JAEA-Research 2025-002, 18 Pages, 2025/05
It is advocated as a development target of fast reactors (FRs) to allow for the of use of mixed oxide (MOX) fuels containing minor actinide (MA) separated and recovered from spent fuels with the aim of reducing the volume and toxicity of high-level radioactive waste generated from nuclear reactors. In the development of MAMOX fuels, it is important behavior to understand the thermal properties such as thermal conductivity for fuel design and analysis of the irradiation. However, there are only a few reports on the thermal properties of MA-MOX fuels, and neither the effects of MA contents nor of oxygen non-stoichiometry in MOX fuels on their thermal conductivities have been fully understood. In this study, the thermal conductivities of MOX fuels with up to 15% Am content were measured at near-stoichiometric composition and the relationship between thermal conductivity and Am content was evaluated. Moreover, the thermal conductivities of Am-doped UO fuels were also measured and evaluated by comparison with Am-MOX to evaluate the effect of Am content. The fuel samples used in this study were three types of MOX with a Pu content of 30% and different Am contents (5%, 10%, and 15%), and UO
containing 15% Am. The thermal conductivities of specimens were calculated from the thermal diffusivities measured by the laser flash method, the density of the specimens and, the heat capacity at constant pressure. The oxygen partial pressure during the measurement was controlled at that of the targeted near-stoichiometric composition. The thermal conductivities of all specimens exhibited a decline with increasing temperature and Am content, with a particularly pronounced reduction observed below 1,173 K. The results of the classical phonon scattering model analysis of the measured thermal conductivities showed that the effect of lattice strain due to the Am addition was significant on the thermal resistivity change, and the effect was comparable for both MOX and UO
.
Yokoyama, Keisuke; Watanabe, Masashi; Usui, Akane; Seki, Takayuki*; Onishi, Takashi; Kato, Masato
Nuclear Materials and Energy (Internet), 42, p.101908_1 - 101908_6, 2025/03
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Oxygen potential of high Am content MOX, (UPu
Am
)O
, was measured at 1273 K, 1473 K, 1573 K, and 1623 K. by gas equilibrium method using thermogravimeter. Comparing the measured data with the literature data, it was found that the addition of 15% Am increases the oxygen potential of (U, Pu)O
by 100-150 kJ/mol for the same Pu content and O/M ratio. The proportion of cations in the stoichiometric composition was determined as (U
U
Pu
Am
)O
, assuming the presence of Am
and partial oxidation of U
to U
. The relationship between oxygen partial pressure and deviation x from stoichiometry in (U
Pu
Am
)O
was analyzed by defect chemistry model. The equation to represent the O/M ratio was derived as a function of temperature and oxygen partial pressure. A part of this study includes the results of MEXT Innovative Nuclear Research and Development Program Grant Number JPMXD0219214921.
Fujita, Tatsuya; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 62(2), p.179 - 196, 2025/02
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)This study newly established a direct coupling code system consisting of the nuclear data processing code FRENDY version 2, and the three-dimensional heterogeneous transport code GENESIS (FRENDY-V2/GENESIS) for easy implementation of the random-sampling-based uncertainty quantification considering the implicit effect due to nuclear cross-section (XS) perturbations. The multi-group macroscopic XSs prepared for GENESIS were generated by FRENDY version 2, where the Dancoff factor was calculated by the neutron current method. Then the background XSs were evaluated based on the Carlvik two-term rational approximation. The infinite multiplication factor (k-infinity) and the fission reaction rate distribution in UO and MOX lattice geometries were compared with MVP3 to verify the calculation accuracy of FRENDY-V2/GENESIS. The sensitivity analyses on the discretization conditions such as the ray tracing of the method of characteristics were also carried out. Through several comparisons between FRENDY-V2/GENESIS and MVP3, FRENDY-V2/GENESIS with the SHEM 361-group structure calculates the k-infinity within approximately 50 pcm and the fission reaction rate distribution within approximately 0.1% by the root mean square, respectively. Consequently, the applicability of FRENDY-V2/GENESIS was verified, and FRENDY-V2/GENESIS can be used to discuss the implicit effect due to multi-group XS perturbations.
Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 199(1), p.18 - 41, 2025/01
Times Cited Count:1 Percentile:57.00(Nuclear Science & Technology)A series of integral experiments were conducted at FCA of JAEA, simulating LWR cores with a tight lattice cell of highly enriched MOX fuel containing more than 15% fissile plutonium. The three experimental configurations were constructed using foamed polystyrene with different void fractions to clarify the prediction accuracy of neutronic calculation codes and nuclear data among various neutron spectra. The nuclear characteristics measured in the experiments were criticality, moderator void reactivity worths, and sample reactivity worths. The preliminary analyses on experiments were conducted using a deterministic calculation code conventionally used for fast reactors with JENDL-4.0. Most reactivity worth calculations correlated well with the experimental values. Specifically for the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations were performed with MVP3 code by modeling the experimental setup in detail, confirming the validity of the deterministic calculations.
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 UPu
O
(y = 0.30 and 0.45) and PuO
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
50 K lower than previously reported. Finally, the systematic experimental procedure allowed us refining the locus of the solvus existing in hypostoichiometric U
0Pu
O
, U
Pu
O
, and PuO
dioxides.
Hirooka, Shun; Morimoto, Kyoichi; Matsumoto, Taku; Ogasawara, Masahiro*; Kato, Masato; Murakami, Tatsutoshi
Journal of Nuclear Materials, 598, p.155188_1 - 155188_9, 2024/09
Times Cited Count:0 Percentile:0.00(Materials Science, Multidisciplinary)no abstracts in English
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:57.00(Materials Science, Multidisciplinary)Frazer, D.*; Saleh, T. A.*; Matsumoto, Taku; Hirooka, Shun; Kato, Masato; McClellan, K.*; White, J. T.*
Nuclear Engineering and Design, 423, p.113136_1 - 113136_7, 2024/07
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Nanoindentation based techniques can be employed on minute volumes of material to measure mechanical properties, including Young's modulus, hardness, and creep stress exponents. In this study, (U,Ce)O solid solutions samples are used to develop elevated temperature nanoindentation and nanoindentation creep testing methods for use on mixed oxide fuels. Nanoindentation testing was performed on 3 separate (Ux-1,Cex)O
compounds ranging from x equals 0.1 to 0.3 at up to 800
C: their Young's modulus, hardness, and creep stress exponents were evaluated. The Young's modulus decreases in the expected linear manner while the hardness decreases in the expected exponential manner. The nanoindentation creep experiments at 800
C give stress exponent values, n=4.7-6.9, that suggests dislocation motion as the deformation mechanism.
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:25.62(Nuclear Science & Technology)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:4 Percentile:34.13(Materials Science, Ceramics)Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka
Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 15 Pages, 2024/00
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Hirooka, Shun; Horii, Yuta; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*; Vauchy, R.; Hayashizaki, Kohei; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato
Journal of Nuclear Science and Technology, 60(11), p.1313 - 1323, 2023/11
Times Cited Count:5 Percentile:80.64(Nuclear Science & Technology)Additive MOX pellets are fabricated by a conventional dry powder metallurgy method. NdO
and Sm
O
are chosen as the additive materials to simulate the corresponding soluble fission products dispersed in MOX. Shrinkage curves of the MOX pellets are obtained by dilatometry, which reveal that the sintering temperature is shifted toward a value higher than that of the respective regular MOX. The additives, however, promote grain growth and densification, which can be explained by the effect of oxidized uranium cations covering to a pentavalent state. Ceramography reveals large agglomerates after sintering, and Electron Probe Micro-Analysis confirms that inhomogeneous elemental distribution, whereas XRD reveals a single face-centered cubic phase. Finally, by grinding and re-sintering the specimens, the cation distribution homogeneity is significantly improved, which can simulate spent nuclear fuels with soluble fission products.
Yokoyama, Keisuke; Uwaba, Tomoyuki
Journal of Nuclear Science and Technology, 60(10), p.1219 - 1227, 2023/10
Times Cited Count:2 Percentile:46.61(Nuclear Science & Technology)no abstracts in English
Ikusawa, Yoshihisa; Nagayama, Masahiro*
JAEA-Data/Code 2023-006, 24 Pages, 2023/07
Core fuels with stainless steel cladding and high plutonium content mixed oxide (MOX) fuel in a water-cooled environment, such as supercritical water-cooled reactors (SCWR) and reduced-moderation water reactors (RMWR), have been studied. In order to contribute to the research and development of such a core fuel concept, the fuel performance code "FEMAXI-8" was verified based on the results of post irradiation examinations of MOX fuel irradiated in the experimental fast reactor "JOYO". FEMAXI-8 is the latest version of the behavior analysis code developed by JAEA to analyze the behavior of light water reactor fuels under normal operation and transient conditions. This latest code has been improved and developed to allow the selection of stainless steel cladding property models to analyze improved fuels such as accident tolerant fuels. The purpose of this report is to confirm the prediction accuracy of FEMAXI-8 for the irradiation behavior of the new type of core fuel that is currently being developed. As a result of the verification, it was confirmed that FEMAXI-8 has sufficient analysis accuracy for the irradiation behavior of sodium-cooled fast reactor MOX fuel with stainless steel cladding, which exceeds the plutonium content and irradiation conditions of light water reactors. In the future, the analysis accuracy of FEMAXI-8 could be improved by adopting the O/M ratio dependence of MOX fuel thermal conductivity and the irradiation behavior evaluation model at high temperature.
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:94.10(Materials Science, Multidisciplinary)Minari, Eriko*; Kabasawa, Satsuki; Mihara, Morihiro; Makino, Hitoshi; Asano, Hidekazu*; Nakase, Masahiko*; Takeshita, Kenji*
Journal of Nuclear Science and Technology, 60(7), p.793 - 803, 2023/07
Times Cited Count:3 Percentile:42.88(Nuclear Science & Technology)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:14.76(Nuclear Science & Technology)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 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.
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.