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Yamaji, Akifumi*; Susuki, Naomichi*; Kaji, Yoshiyuki
IAEA-TECDOC-1921, p.199 - 209, 2020/07
The thermo-physical models and irradiation behavior of FeCrAl as defined by the benchmark organizer have been implemented to FEMAXI-7. Analyses were carried out firstly for the specified normal operation condition. Then, some sensitivity analyses were carried out with different assumptions and model parameters. Under the normal operating condition, the predicted FeCrAl cladded fuel performance was similar to that of Zry cladded fuel with notable, but not major difference regarding late gap closure. Under the simulated LOCA conditions, the burst pressure could be evaluated. The predicted cladding creep strain at burst was mainly attributed to creep strain with negligible plastic strain. Overall, FEMAXI-7 analyses have demonstrated excellent robustness and flexibility in modeling FeCrAl-UO
system under normal and LOCA conditions.
Tanaka, Masaaki; Nagasawa, Kazuyoshi*
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.6650 - 6663, 2015/08
For the fundamental validation of fluid-structure thermal interaction code (MUGTHES), numerical simulations for the planar triple parallel jets tests in WAJECO and PLAJEST have been conducted as the benchmark analysis. In comparison between the numerical results and the provided experimental results, thermal mixing process and large-scale eddy structures generated in the triple jets mixing and the relation between temperature fluctuation generation and large-eddy structures were revealed. And also, the attenuation process of temperature fluctuation from the fluid to the structure was indicated.
PNC TN9410 93-037, 99 Pages, 1992/12
In the frame work of international cooperation between PNC and European AGT9B, benchmark studies were planned on the structural integrity analyses, Among these, a problem on the dynamic buckling of a hemisphere was supplied by the European side. A small annular gap between two concentric hemisphere is filled with water and the whole system is subjected to a vertical harmonic excitation so that the inner thin shell buckles by the dynamic fluid pressure. The problem is to infer the frequency of the excitation and the pressure at which the hemisphere buckles. An intensive series of analyses were performed using a general purpose non-linear finite element code, FINAS. The analyses were a blind test since no information was available on the experimental results. It is inferred as a result of the analyses that; (1)The frequency of the excitation is 27 Hz. (2)The critical buckling pressure is about 0.16 MPa. (3)The fundamental vibration mode and the buckling mode are axi-symmetric, and the buckling occurs in the elastic range. Discussion on the adequacy of the analyses will be made when the experimental result is available.
*; *; Shin, Kazuo*; Masukawa, Fumihiro; Naito, Yoshitaka
JAERI-M 91-013, 54 Pages, 1991/02
no abstracts in English
*;
JAERI-M 7067, 61 Pages, 1977/04
no abstracts in English
Masaki, Koichi; Uno, Shumpei*; Katsuyama, Jinya; Li, Y.
no journal, ,
no abstracts in English
Horiguchi, Naoki; Nakamura, Koichi*; Sekiguchi, Takashige*; Uesawa, Shinichiro; Himi, Masashi*; Yoshida, Hiroyuki; Nishimura, Satoshi*
no journal, ,
no abstracts in English
Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki
no journal, ,
Validation of an analysis model for a plant dynamic analysis code named Super-COPD including neutronics calculation of a one-point reactor kinetics model necessitates the further work on the beyond design basis accident. Therefore, JAEA participated in IAEA benchmark for Loss of Flow without Scram (LOFWOS) test No.13 performed at the Fast Flux Test Facility (FFTF), and the transient analysis at the first blind phase considering with major reactivity feedback mechanisms was carried out. It was observed that the whole plant dynamics analysis could follow the measured data. As a future work, the gap conductance model for transient, the upper plenum of reactor vessel with dividing several regions or multi-dimensional modeling, and the core model that can evaluate the radial heat transfer rate more accurately will be refined.
Nobuhara, Fumiyoshi*; Matsuda, Norihiro; Onishi, Seiki*; Matsui, Yusuke*; Kubota, Osamu*; Sakamoto, Yukio*; Hirao, Yoshihiro*
no journal, ,
no abstracts in English
