Safety analyses of Kaist Micro modular reactor with modified Gamma+ code

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To supply distributed power system to remote region, a concept of micro reactor with supercritical CO2 (S-CO2) cooled direct Brayton cycle called KAIST Micro Modular Reactor, MMR in short, has been developed. The MMR has a small reactor core with power output of 36MWth, power conversion system and passive decay heat removal system in a double wall containment whose weight is approximately 150 tons and dimensions are 7.0 m in length and 3.8 m in diameter. The dimension and weight are sized so that the single modular of whole MMR can be transported by a ship or a truck. Design parameters and configuration of the reactor core, power conversion system and passive decay heat removal system have been optimized by the KAIST-research team. Until now, only the on-design performances were obtained. The nuclear system design can be finalized after postulated accidents are simulated to guarantee its safety. The S-CO2 system code platform is prepared to simulate various transient conditions of the S-CO2 Brayton cycle. GAMMA+ code, which is developed by KAERI for gas-cooled reactor system analysis, is modified for the S-CO2 system analysis. Postulated accidents of MMR are also simulated to assure whether its integrity is maintained during the accidents. Among various postulated accidents, loss of external load (LOL) is first modeled because MMR will be operating in a remote region where grid infrastructure isn’t well developed. Furthermore, since MMR has a high pressure boundary, loss of coolant accidents (LOCA) are also analyzed. While both LOL and LOCA are initiated, it is assumed that the reactor shutdown system is not available and single failure of safety features is assumed for conservative assessment. It is concluded that the current design of MMR has an ability to keep its integrity for the analyzed accident scenarios.
Publisher
Association for Computing Machinery, Inc
Issue Date
2017-09
Language
English
Citation

17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017

URI
http://hdl.handle.net/10203/310455
Appears in Collection
NE-Conference Papers(학술회의논문)
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