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Horii, Yuta; Hirooka, Shun; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*; Furusawa, Naoya*; Murakami, Tatsutoshi; Kato, Masato
Journal of Nuclear Materials, 588, p.154799_1 - 154799_20, 2024/01
Times Cited Count：7 Percentile：85.99(Materials Science, Multidisciplinary)The thermal conductivities of near-stoichiometric (U,Pu,Am)O
doped with NdO
/SmO, which is major fission product (FP) generated by a uranium-plutonium mixed oxides (MOX) fuel irradiation, as simulated fission products are evaluated at 1073-1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneity of Nd/Sm are prepared using two kinds of powder made by ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the classical phonon transport model (A+BT)
. The dependences of the coefficients A and B on the Nd/Sm content (C
and C
, respectively) are evaluated as: A(mK/W)=1.70 
10
+ 0.93C + 1.20C, B(m/W)=2.39 10
.
Bess, J. D.*; Chipman, A. S.*; Pope, C. L.*; Jensen, C. B.*; Ozawa, Takayuki; Hirooka, Shun; Kato, Masato*
Nuclear Science and Engineering, 197(8), p.1845 - 1872, 2023/08
Pretransient characterization was performed for the EBR-II MOX fuel pellets from the SPA-2/-2B Operational Reliability Testing collaboration between Japan and US. The continued collaboration will investigate the transient performance of these rods in TREAT at Idaho National Laboratory. The results will fill a gap in existing transient performance data for MOX as these rods have a peak burnup of ~134.4 GWd/t in the EBR-II. Fuel pellet properties were gathered from available resources and their irradiation and decay history evaluated. Further reactor physics calculations were performed to support the experiment design, reactor operations, and safety analyses necessary to enable the programmatic success. Of the three irradiated fuel pins, two will undergo transient testing, and all three will undergo post-irradiation examination.
Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi
Nuclear Technology, 207(8), p.1280 - 1289, 2021/08
Times Cited Count：3 Percentile：27.12(Nuclear Science & Technology)Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.
Kakiuchi, Kazuo; Udagawa, Yutaka; Amaya, Masaki
Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06
Times Cited Count：2 Percentile：20.79(Nuclear Science & Technology)Udagawa, Yutaka; Fuketa, Toyoshi*
Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08
Uwaba, Tomoyuki; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*
Nuclear Engineering and Design, 331, p.186 - 193, 2018/05
Times Cited Count：5 Percentile：40.11(Nuclear Science & Technology)A computer code for the analysis of the overall irradiation performance of a fast reactor mixed-oxide (MOX) fuel element was coupled with a specialized code for the analysis of fission product cesium behaviors in a MOX fuel element. The coupled code system allowed for the analysis of the radial and axial Cs migrations, the generation of Cs chemical compounds and fuel swelling due to Cs-fuel-reactions in association with the thermal and mechanical behaviors of the fuel element. The coupled code analysis was applied to the irradiation performance of a fast reactor MOX fuel element attaining high burnup for discussion on the axial distribution of Cs, fuel-to-cladding mechanical interaction owing to the Cs-fuel-reactions by comparing the calculated results with post irradiation examinations.
; ;
Journal of Nuclear Science and Technology, 25(5), p.456 - 463, 1988/05
Times Cited Count：4 Percentile：46.39(Nuclear Science & Technology)no abstracts in English
Inabe, Teruo; ;
JAERI-M 9178, 23 Pages, 1980/11
no abstracts in English
Riyana, E. S.; Okumura, Keisuke; Sakamoto, Masahiro; Matsumura, Taichi; Terashima, Kenichi
no journal, ,
