A preliminary study on seismic analysis of graphite cores for gas-cooled microreactor
<bold>Background</bold>Graphite serves as a vital structural material in high-temperature gas-cooled reactors. Ensuring its integrity under seismic loading is crucial for maintaining core safety.<bold>Purpose</bold>This study aims to assess the seismic response of a gas-cooled microreactor (GCMR) with graphite core and validate its compliance with the ASME design code.<bold>Methods</bold>GCMR with horizontal stacking structure of graphite components connected by pin keys was taken as seismic analysis object. Experimental collision tests and simulations involving small-scale graphite blocks were performed to determine stiffness and damping parameters. The finite element simulation calculation was carried out for several working conditions with oblique collision angle in the range of 0.01°~0.1°. Subsequently, a simplified core model incorporating equivalent spring-damper elements was developed and analyzed using time-history simulations to obtain the interaction force between graphite components, and the structure of GCMR with graphite core was evaluated by the failure probability method.<bold>Results</bold>Simulation and test results show that the Kelvin collision model successfully characterizes interactions between graphite blocks, exhibiting stiffness values ranging from 1.70×10<sup>9</sup> N∙m<sup>-1</sup> to 2.20×10<sup>9</sup> N∙m<sup>-1</sup> for collision angles of 0.02°~0.05°. The peak collision force between graphite components reaches 1.21×10<sup>5</sup> N. Stress assessment based on the Weibull distribution reveals a failure probability of less than 0.5% under design-level seismic loads.<bold>Conclusions</bold>The stiffness-equivalent method proposed in this study offers an effective means of simulating nonlinear core behavior, affirming the structural safety of the graphite core under seismic conditions.
Authors
- Tianyou Feng (ORCID: https://orcid.org/0000-0003-1024-7704)
- Tianbao Lan (ORCID: https://orcid.org/0000-0003-1847-2463)
- Zhong Huang (ORCID: https://orcid.org/0000-0002-6580-3870)
- Sheng Feng
Publication Details
- Journal
- Nuclear Techniques
- Published
- 2025-10-01
- DOI
- https://doi.org/10.3724/j.0253-3219.2025.hjs.48.250014
- Citations
- 1
- Primary Topic
- Graphite, nuclear technology, radiation studies
- Type
- article
- Field-Weighted Citation Impact
- 0.52
Funders
- China National Nuclear Corporation