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Nishi, Yoshihisa*; Ueda, Nobuyuki*; Kinoshita, Izumi*; Miyakawa, Akira; Kato, Mitsuya*
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 10 Pages, 2006/07
CERES is plant system analysis code for LMRs developed by the Central Research Institute of Electric Power Industry (CRIEPI). CERES has a function of calculating multidimensional flow in the plena of a coolant in addition to that in one-dimensional plant network calculation. To verify the CERES code, analyses were performed by using the result of the plant trip test from the partial power operation of the prototype FBR "MONJU" that had been executed in December, 1995. The verification work was performed as a joint research of CRIEPI and JAEA. (1)Analysis concerning the primary/secondary/auxiliary cooling system (the plenum in the reactor vessel (R/V) is modeled in R-Z 2-dimension). (2)Analysis concerning the flow pattern in the plenum of R/V (the plenum is modeled in 3-dimension). (3)Analysis concerning the flow pattern inside the IHX plenum (the plenum in the IHX is modeled in 3-dimension). Analytical results by the CERES code showed good agreement with the results of the test of the "MONJU". Fundamental abilities of the CERES as a plant dynamics calculation code had been verified through these analyses. Additionally, some characteristic flows in plena of "MONJU" became clear by these analyses.
Nishi, Yoshihisa*; Ueda, Nobuyuki*; Kinoshita, Izumi*; Miyakawa, Akira; Kato, Mitsuya*
JNC TY2400 2005-001, 66 Pages, 2005/06
Multi-dimensional thermal-hydraulic characteristic of the coolant in the reactor vessel (R/V) influences the temperature at the plant transitional condition of fast breeder reactor (FBR). CRIEPI is developing plant dynamics calculation code CERES for FBR that adds multi-dimensional thermal-hydraulic analysis function to one-dimensional system calculation code to evaluate the temperature distribution in high accuracy. The temperature distribution affects the integrity of equipments of FBR. To verify the CERES code, analyses were performed by using the result of the plant trip test from the partial power operation of the prototype fast breeder reactor
Yamada, Fumiaki; Hashimoto, Akihiko*; Kato, Mitsuya*; Arikawa, Mitsuhiro*
no journal, ,
In this report that review on the safety evaluation for the consequence of Large Pipe Break in Primary Heat Transport System on the Monju used experimental data.
Kato, Mitsuya*; Takano, Masahito*; Morizono, Koji
no journal, ,
no abstracts in English
Mori, Takero; Araki, Kosuke*; Kato, Mitsuya*; Takano, Masahito*
no journal, ,
An improved analysis model and the actual component characteristic data for the main cooling system are incorporated in Monju plant dynamics analysis code: Super-COPD. The verification of this new analysis model is based on the results of a plant trip test at 40% rated power and on plant control system characteristics.
Miyagawa, Takayuki; Kitano, Akihiro; Muranaka, Makoto; Kato, Mitsuya*; Okawachi, Yasushi
no journal, ,
no abstracts in English
Miyagawa, Takayuki; Kitano, Akihiro; Muranaka, Makoto; Kato, Mitsuya*; Okawachi, Yasushi
no journal, ,
no abstracts in English