The Supercritical Carbon Dioxide (S-CO2) Brayton Cycle is a promising advanced alternative to the Rankine saturated steam cycle and recuperated gas Brayton cycle for the energy converters of specific reactor concepts belonging to the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. A new plant dynamics analysis computer code has been developed for simulation of the S-CO2 Brayton cycle coupled to an autonomous, natural circulation Lead-Cooled Fast Reactor (LFR). The plant dynamics code was used to simulate the whole-plant response to accident conditions. The specific design features of the reactor concept influencing passive safety are discussed and accident scenarios are identified for analysis. Results of calculations of the whole-plant response to loss-of-heat sink, loss-of-load, and pipe break accidents are demonstrated. The passive safety performance of the reactor concept is confirmed by the results of the plant dynamics code calculations for the selected accident scenarios.
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14th International Conference on Nuclear Engineering
July 17–20, 2006
Miami, Florida, USA
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
0-7918-4244-4
PROCEEDINGS PAPER
Transient Accident Analysis of a Supercritical Carbon Dioxide Brayton Cycle Energy Converter Coupled to an Autonomous Lead-Cooled Fast Reactor
Anton Moisseytsev,
Anton Moisseytsev
Argonne National Laboratory, Argonne, IL
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James J. Sienicki
James J. Sienicki
Argonne National Laboratory, Argonne, IL
Search for other works by this author on:
Anton Moisseytsev
Argonne National Laboratory, Argonne, IL
James J. Sienicki
Argonne National Laboratory, Argonne, IL
Paper No:
ICONE14-89544, pp. 623-634; 12 pages
Published Online:
September 17, 2008
Citation
Moisseytsev, A, & Sienicki, JJ. "Transient Accident Analysis of a Supercritical Carbon Dioxide Brayton Cycle Energy Converter Coupled to an Autonomous Lead-Cooled Fast Reactor." Proceedings of the 14th International Conference on Nuclear Engineering. Volume 3: Structural Integrity; Nuclear Engineering Advances; Next Generation Systems; Near Term Deployment and Promotion of Nuclear Energy. Miami, Florida, USA. July 17–20, 2006. pp. 623-634. ASME. https://doi.org/10.1115/ICONE14-89544
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