Abstract
One of the main advantages of the continuous-energy Monte Carlo method is its versatility and the capability to model any fuel or reactor configuration without major approximations. This capability becomes particularly valuable in studies involving innovative reactor designs and next-generation systems, which often lie beyond the capabilities of deterministic light water reactor transport codes. In this study, a conceptual prismatic high-temperature gas-cooled reactor (HTGR) fuel assembly was modeled using the Serpent Monte Carlo reactor physics burnup calculation code, under development at VTT Technical Research Centre of Finland since 2004. A new geometry model was developed for the Serpent code in order to account for the heterogeneity effects encountered in randomly dispersed particle fuels. The results are compared to other Monte Carlo and deterministic transport codes, and the study also serves as a test case for the modules and methods in SCALE 6.
| Original language | English |
|---|---|
| Pages (from-to) | 782-784 |
| Number of pages | 3 |
| Journal | Transactions of the American Nuclear Society |
| Volume | 101 |
| Publication status | Published - 2009 |
| MoE publication type | A1 Journal article-refereed |
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Leppänen, J., & DeHart, M. (2009). HTGR Reactor Physics and Burnup Calculations Using the Serpent Monte Carlo Code. Transactions of the American Nuclear Society, 101, 782-784.