Level 1 probabilistic safety assessment of supercritical–CO2–cooled micro modular reactor in conceptual design phaseopen accessLevel 1 probabilistic safety assessment of supercritical – CO 2 – cooled micro modular reactor in conceptual design phase
- Authors
- So, E.; Kim, M.C.
- Issue Date
- Feb-2021
- Publisher
- Korean Nuclear Society
- Keywords
- Advanced reactor; Event tree analysis; Fault tree analysis; Initiating event category; Passive safety system; Probabilistic safety assessment
- Citation
- Nuclear Engineering and Technology, v.53, no.2, pp 498 - 508
- Pages
- 11
- Journal Title
- Nuclear Engineering and Technology
- Volume
- 53
- Number
- 2
- Start Page
- 498
- End Page
- 508
- URI
- https://scholarworks.bwise.kr/cau/handle/2019.sw.cau/52669
- DOI
- 10.1016/j.net.2020.07.029
- ISSN
- 1738-5733
- Abstract
- Micro reactors are increasingly being considered for utilization as distributed power sources. Hence, the probabilistic safety assessment (PSA) of a direct supercritical–CO2–cooled fast reactor, called micro modular reactor (MMR), was performed in this study; this reactor was developed using innovative design concepts. It adopted a modular design and passive safety systems to minimize site constraints. As the MMR is in its conceptual design phase, design weaknesses and valuable safety insights could be identified during PSA. Level 1 internal event PSA was carried out involving literature survey, system characterization, identification of initiating events, transient analyses, development of event trees and fault trees, and quantification. The initiating events and scenarios significantly contributing to core damage frequency (CDF) were determined to identify design weaknesses in MMR. The most significant initiating event category contributing to CDF was the transients with the power conversion system initially available category, owing to its relatively high occurrence frequency. Further, an importance analysis revealed that the safety of MMR can be significantly improved by improving the reliability of reactor trip and passive decay heat removal system operation. The findings presented in this paper are expected to contribute toward future applications of PSA for assessing unconventional nuclear reactors in their conceptual design phases. © 2020
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