HTGR Reactor Physics and Burnup Calculations Using the Serpent Monte Carlo Code
- VTT TECHNICAL RESEARCH CENTRE OF FINLAND
- ORNL
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 LWR transport codes. In this study, a conceptual prismatic 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 explicit particle fuel model was developed to account for the heterogeneity effects. 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.
- Research Organization:
- Oak Ridge National Laboratory (ORNL)
- Sponsoring Organization:
- ORNL work for others
- DOE Contract Number:
- AC05-00OR22725
- OSTI ID:
- 980700
- Country of Publication:
- United States
- Language:
- English
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