The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in over 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.

The Serpent Monte Carlo code: Status, development and applications in 2013

Accelerator driven systems for transmutation: Fuel development, design and safety

Neutronic performance of uranium nitride composite fuels in a PWR

Enhanced thermal conductivity accident tolerant fuels for improved reactor safety – A comprehensive review

Screening of advanced cladding materials and UN–U3Si5 fuel

Neutronic evaluation of a PWR with fully ceramic microencapsulated fuel. Part I: Lattice benchmarking, cycle length, and reactivity coefficients

Neutronic evaluation of a PWR with fully ceramic microencapsulated fuel. Part II: Nodal core calculations and preliminary study of thermal hydraulic feedback

Examining the inherent safety of prism, SAFR, and the MHTGR

This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

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A.L. Aronson

Papers

Neutronic performance of uranium nitride composite fuels in a PWR

Neutronic evaluation of a PWR with fully ceramic microencapsulated fuel. Part I: Lattice benchmarking, cycle length, and reactivity coefficients

Neutronic evaluation of a PWR with fully ceramic microencapsulated fuel. Part II: Nodal core calculations and preliminary study of thermal hydraulic feedback

Examining the inherent safety of prism, SAFR, and the MHTGR

Intercomparison of results for a pwr rod ejection accident