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Kamide, Hideki; *; Kimura, Nobuyuki;
JNC TN9400 2001-017, 86 Pages, 2000/09
JNC-TN9400-2001-017.pdf:2.66MB
Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. Flow rate through the reactor core depends on temperature distribution in the cooling systems under natural circulation condition. The flow rate and temperature distribution couples with each other and thermohydraulic phenomena have complex relations of cause and effect. A numerical estimation method of such natural circulation phenomena in a reactor core has been developed, which is based on the multi-dimensional thermohydraulic analysis code, AQUA. The prediction method models each subassembly as a rectangular duct and gap region between the subassemblies with the upper plenum in a reactor vessel. This modeling enables us to calculate the inter-wrapper flow (IWF, gap flow between the subassemblies), thermal interaction between IWF and flow in the subassemblies, and inter-subassembly heat transfer, and flow redistribution in the core. This numerical simulation method was verified based on experimental data of a sodium test using 7-subassembly core model and also a water test which simulates inter-wrapper flow using the 1/12 sector model of a large scale reactor core. We applied the estimation method to the natural circulation phenomena in a large scale fast reactor of 600 MWe. It was shown that IWF depended on pad geometry of the subassembly and cooled the reactor core effectively. The highest temperature was shown in a outer core subassembly. The temperature in the core depends on IWF, flow redistribution in the core, and inter-subassembly heat transfer.
; *; ; Kamide, Hideki
JNC TN9410 99-004, 66 Pages, 1999/01
Instability analysis was carried out using BOST code for a steam generator in a large scale sodium test facility of fast reactors. However, it was found that BOST code gave stable characteristics under the conditions of higher pressure in water-steam system than MONJU conditions, even if the flow ratio of sodium to water was increased as expected to give unstable condition. Here, modification of BOST code was considered and we found some points to be modified. However, main reasons of stable calculation were not resolved. In this report, the current status of BOST code was summarized especially for the stable calculation under the higher pressure condition for further modification and a new code based on current knowledge and coding technique.
; *; ; ;
PNC TN9410 91-109, 83 Pages, 1991/02
Many studies for the 600MWe large FBR plant design began at April 1990 in our section. In order to evaluate the thermal transient for this plant, the analysis code, by which we can easily find the proper design parameters, is under development. The program of this code is based on Super-COPD. As the first step of the development, a simple model of Super-COPD was constructed, and the thermal transients of this plant were calculated. These results showed as follows; (1)The thermal transient characteristics were easily calculated when the primary and secondary flows were input conditions, (2)The results of this calculation showed the similar characteristics to those of 'Monju' plant except for the AC outlet sodium temperature that changes much in small differences in its flow characteristics, (3)The thermal transients of this large FBR plant transport the response much quicker than those of 'Monju'. It is supposed that this plant has shorter loops than 'Monju'. Another calculations of this plant and 'Monju' plant are scheduled in order to verify this model.