RSICC is authorized to distribute SERPENT 1.1.7 for research and education purposes only. Commercial use is prohibited.

SERPENT 1.1.7: Continuous Energy Monte Carlo Reactor Physics Burnup Calculation Code.

**RESTRICTIONS**

RSICC is authorized to distribute SERPENT 1.1.7 for research and education purposes only. Commercial use is prohibited.

**DATA
LIBRARIES **

Serpent 1.1.7 cross section library based on ENDF/B-VI.8

Serpent 1.1.7 cross section library based on ENDF/B-VII

Serpent 1.1.7 cross section library based on JEF-2.2

Serpent 1.1.7 cross section library based on JEFF-3.1.1

Serpent 1.1.7 thermal scattering libraries based on JEF-2.2, JEFF3.1, ENDF/B-VI.8 and ENDF/B-VII

VTT Technical Research Centre of Finland.

ANSI-C; Linux-based PC, Macintosh, UNIX workstations (RSICC ID: C00757MNYWS00).

SERPENT is a three-dimensional, continuous-energy Monte Carlo reactor physics burnup calculation code specifically designed for lattice physics applications. The code uses built-in calculation routines for generating homogenized multi-group constants for deterministic reactor simulator calculations. The standard output includes effective and infinite multiplication factors, homogenized reaction cross sections, scattering matrices, diffusion coefficients, assembly discontinuity factors, point-kinetic parameters, effective delayed neutron fractions, and precursor group decay constants. User-defined tallies can be set up for calculating various integral reaction rates and spectral quantities.

Internal burnup calculation capability allows SERPENT to simulate fuel depletion as a completely stand-alone application. Extensive effort has been put into optimizing the calculation routines and the code is capable of running detailed assembly burnup calculations similar to deterministic lattice codes within a reasonable calculation time. The overall running time can be further reduced by parallelization.

SERPENT can be used for various reactor physics calculations at pin, assembly and core levels. The continuous-energy Monte Carlo method allows the modeling of any critical reactor type, including both thermal and fast neutron systems. The suggested applications of SERPENT include group constant generation, fuel cycle studies, validation of deterministic lattice physics codes, and educational, training and demonstration purposes.

A complete description of the project is found at the SERPENT website - http://montecarlo.vtt.fi.

SERPENT uses the continuous-energy Monte Carlo criticality source method for simulating neutron transport in a self-sustaining system. Cross sections are read from ACE format data libraries and reconstructed on a single unionized energy grid to speed up the calculation. Interaction physics is based on classical collision kinematics and ENDF reaction laws.

The geometry description follows the standard Monte Carlo approach based on universes, cells and surfaces, which allow the modeling of practically any two- or three-dimensional fuel or reactor configuration. The tracking routine uses conventional surface-to-surface ray-tracing with the Woodcock delta-tracking method. The combination of the two methods has been found efficient and well-suited for lattice physics applications.

The Bateman depletion equations in the burnup calculation mode are solved using either the Transmutation Trajectory Analysis Method (TTA) or a matrix exponential solution based on the Chebyshev Rational Approximation Method (CRAM). Radioactive decay and fission yield data is read from standard ENDF format data files.

Parallel calculation mode is available using the Message Passing Interface (MPI).

Minor code updates are distributed by the development team to registered users by email. Registration can be accomplished by contacting the development team.

The running time depends on the case and the calculation parameters. Two-dimensional infinite-lattice calculations involving 3 million neutron histories usually take some 5 to 20 minutes on a single-processor 3 GHz PC workstation. A test LWR assembly burnup calculation with more than 250 actinide and fission product nuclides, 65 separate depletion zones, 40 burnup steps with predictor-corrector calculation and 3 million neutron histories per transport cycle was completed in 15 hours on the same computer.

SERPENT runs on standard Linux, Mac or UNIX Systems. Memory demand may become a limiting factor in burnup calculation. At least 5 GB of RAM is recommended if the code is intended to be used for assembly burnup calculations.

SERPENT has been developed under PC Linux and Mac OS X operating systems. A standard C-compiler (gcc) is needed for building the source code. MPI libraries must be installed to run SERPENT in the parallel calculation mode. The code uses the GD open source graphics library for producing some graphical output. The source code can also be compiled without the MPI and GD functionality.

Jaakko Leppänen, *PSG2 /
Serpent - a Continuous-energy Monte Carlo Reactor Physics Burnup Calculation
Code Methodology - User's Manual - Validation Report*, VTT Technical Research
Centre of Finland. (September 5, 2009).

PSG2 / Serpent - a Quick Installation Guide.

The package is transmitted as a zip file on a DVD which includes the referenced documents in PDF format, source codes, data files and example problems. No executables are included with the package.

January 2010.

**KEYWORDS**: MONTE CARLO, BURNUP, CONTINUOUS ENERGY,
CRITICALITY, NEUTRON