RSICC CODE PACKAGE CCC-872
RSICC is authorized to distribute SERPENT 2.2.1 for research and education purposes only. Commercial use is prohibited. RSICC can only distribute SERPENT to customers that are in the following “pre-approved” countries: member states of European Union, Australia, South Korea, Canada, Iceland, Japan, New Zealand, Switzerland, the United Kingdom, and the United States.
RSICC can only license and authorize the use of SERPENT for peaceful and legal purposes. SERPENT cannot be used in connection with the development, production, handling, operation, maintenance, storage, detection, identification, or dissemination of chemical, biological, or nuclear weapons or other nuclear explosive devices or the development, production, maintenance, or storage of missiles capable of delivering such weapons. The use of SERPENT for any military purpose is strictly prohibited.
1. NAME AND TITLE
SERPENT 2.2.1: Continuous Energy Monte Carlo Reactor Physics Burnup Calculation Code.
RSICC is authorized to distribute SERPENT 2.2.1 for research and education purposes only. Commercial use is prohibited.
Serpent 2.2.1 cross section library based on ENDF/B-VI.8
Serpent 2.2.1 cross section library based on ENDF/B-VII
Serpent 2.2.1 cross section library based on JEF-2.2
Serpent 2.2.1 cross section library based on JEFF-3.1.1
Serpent 2.2.1 thermal scattering libraries based on JEF-2.2, JEFF3.1, ENDF/B-VI.8 and ENDF/B-VII
VTT Technical Research Centre of Finland.
3. CODING LANGUAGE AND COMPUTER
ANSI-C; Linux-based PC, Macintosh, UNIX workstations (RSICC ID: C00872MNYWS01).
4. NATURE OF PROBLEM SOLVED
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 - https://serpent.vtt.fi/serpent.
5. METHOD OF SOLUTION
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).
6. RESTRICTIONS OR LIMITATIONS
Complex input files may require more compute resources.
7. TYPICAL RUNNING TIME
The running time depends on the case and the calculation parameters.
8. COMPUTER HARDWARE REQUIREMENTS
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.
9. COMPUTER SOFTWARE REQUIREMENTS
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.
Leppänen, J., et al. (2015) "The Serpent Monte Carlo code: Status, development and applications in 2013." Ann. Nucl. Energy, 82 (2015) 142-150..
PSG2 / Serpent - a Quick Installation Guide.
11. CONTENTS OF CODE PACKAGE
The package is transmitted digitally as a zip file that includes source code and documentation. The package includes SERPENT versions 2.2.0 & 2.2.1.
12. DATE OF ABSTRACT
KEYWORDS: MONTE CARLO, BURNUP, CONTINUOUS ENERGY, CRITICALITY, NEUTRON