This paper is a general overview of the Serpent Monte Carlo reactor physics burnup calculation code. The Serpent code is a project carried out at VTT Technical Research Centre of Finland, in an effort to extend the use of the continuous-energy Monte Carlo method to lattice physics applications, including group constant generation for coupled full-core reactor simulator calculations. The main motivation of going from deterministic transport methods to Monte Carlo simulation is the capability to model any fuel or reactor type using the same fundamental neutron interaction data without major approximations. This capability is considered important especially for the development of next-generation reactor technology, which often lies beyond the modeling capabilities of conventional LWR codes. One of the main limiting factors for the Monte Carlo method is still today the prohibitively long computing time, especially in burnup calculation. The Serpent code uses certain dedicated calculation techniques to overcome this limitation. The overall running time is reduced significantly, in some cases by almost two orders of magnitude. The main principles of the calculation methods and the general capabilities of the code are introduced. The results section presents a collection of validation cases in which Serpent calculations are compared to reference MCNP4C and CASMO-4E results.
|Title of host publication||2009 International Nuclear Atlantic Conference - INAC 2009, Rio de Janeiro,RJ, Brazil, September 27 to October 2, 2009|
|Number of pages||11|
|Publication status||Published - 2009|
|MoE publication type||A4 Article in a conference publication|