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Journal Articles

Verification of the plant dynamics analytical code CERES using the results of the plant trip test of the prototype fast breeder reactor MONJU

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.

JAEA Reports

Verification of the plant dynamics analytical code CERES using the results of the plant trip test of the Prototype Fast Breeder Reactor MONJU

Nishi, Yoshihisa*; Ueda, Nobuyuki*; Kinoshita, Izumi*; Miyakawa, Akira; Kato, Mitsuya*

JNC TY2400 2005-001, 66 Pages, 2005/06

JNC-TY2400-2005-001.pdf:7.59MB

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

Oral presentation

Development of Monju plant dynamics analysis code, 2; Improvement of plant initial heat balance settings for Super-COPD

Kato, Mitsuya*; Takano, Masahito*; Morizono, Koji

no journal, , 

no abstracts in English

Oral presentation

Development of Monju plant dynamics analysis code, 3; Verification of Super-COPD code based on the test operation data at 40% rated power

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.

Oral presentation

Prototype FBR Monju system start up test "zero power reactor physics test", 8; Feedback reactivity measurement

Miyagawa, Takayuki; Kitano, Akihiro; Muranaka, Makoto; Kato, Mitsuya*; Okawachi, Yasushi

no journal, , 

no abstracts in English

Oral presentation

Prototype FBR Monju system start up test "Reactor Physics Test", 8; Feedback reactivity measurement

Miyagawa, Takayuki; Kitano, Akihiro; Muranaka, Makoto; Kato, Mitsuya*; Okawachi, Yasushi

no journal, , 

no abstracts in English

Oral presentation

Safety approach to severe accident for new nuclear safety regulation, 7; Review on the safety evaluation for the consequence of large pipe break in PHTS

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.

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