Difference between revisions of "Input syntax manual"
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This page will contain the whole input syntax of Serpent 2, with links to more detailed descriptions where needed. For reference see also the Serpent 1 input manual<ref name="manual">Leppänen, J. | This page will contain the whole input syntax of Serpent 2, with links to more detailed descriptions where needed. For reference see also the Serpent 1 input manual<ref name="manual">Leppänen, J. | ||
| − | + | ''"Serpent – a Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code"'' ([http://montecarlo.vtt.fi/download/Serpent_manual.pdf|User manual], June 18, 2015.</ref> | |
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This is another reference to the same article <ref name="Leppanen1"/> | This is another reference to the same article <ref name="Leppanen1"/> | ||
Revision as of 18:27, 17 November 2015
Contents
Preface
The Serpent code has no interactive user interface. All communication between the code and the user is handled through one or several input files and various output files.
The format of the input file is unrestricted. The file consists of white-space (space, tab or newline) separated words, containing alphanumeric characters(’a-z’, ’A-Z’, ’0-9’, ’.’, ’-’). If special characters or white spaces need to be used within the word (file names, etc.), the entire string must be enclosed within quotation marks.
The input file is divided into separate data blocks, denoted as cards. The file is processed one card at a time and there are no restrictions regarding the order in which the cards should be organized. The input cards are listed below. Additional options are followed by key word "set". All input cards and options are case-insensitive (note to developers: make it so). Each input card is delimited by the beginning of the next card. It is hence important that none of the parameter strings used within the card coincide with the card identifiers.
This page will contain the whole input syntax of Serpent 2, with links to more detailed descriptions where needed. For reference see also the Serpent 1 input manual[1]
This is another reference to the same article [2]
Input cards
mat (material definition)
surf (surface definition)
cell (cell definition)
det (detector definition)
src
Input options
set delnu
set delnu <dnu>
Sets delayed neutron emission on or off. Input values:
| <dnu> | : option to switch delayed neutron emission off (0/no) or on (1/yes) |
Notes:
- Delayed neutron emission is on by default in neutron criticality source and off by default in external source simulations.
set gcut
set gcut <gmax>
Sets generation cut-off for neutrons. Input values:
| <gmax> | : number of simulated generations before cut-off |
The generation cut-off can be used in neutron external source simulations, to limit the length of fission chains.
Notes:
- Applicable only to neutron external source simulation (invoked using (set nps)
- Generation or time cut-off (set tcut) is always needed for neutron external source simulations in super-critical systems.
set nps
set nps <ppb> [ <btch> ]
Sets parameters for simulated particle population in external source mode. Input values:
| <ppb> | : number of particles per batch |
| <btch> | : number of batches |
The simulation is for a number batches using the given batch size.
Notes:
- Using the nps card sets the mode to external source simulation. Criticality source simulation for neutrons is invoked using set pop.
- Running an external source simulation requires a source, defined by the src card. Source definition also sets the transported particle type.
- Neutron external source simulations are limited to sub-critical systems, unless time cut-off (set tcut) or generation cut-off (set gcut) is invoked.
- Neutron external source simulations in multiplying systems may require adjusting the neutron buffer (set nbuf).
- Delayed neutron emission is switched off by default in neutron external source simulation (for compatibility with MCNP). Delayed neutrons can be included with set delnu.
set pop
set pop <npg> <ngen> <nskip> [ <k0> <btch> ]
Sets parameters for simulated neutron population in criticality source mode. Input values:
| <npg> | : number of neutrons per generation |
| <ncyc> | : number of active generations |
| <nskip> | : number of inactive generations |
| <k0> | : initial guess for k-eff |
| <btch> | : batching interval |
The simulation is first run for a number of inactive generations to allow the fission source to converge. This is followed by a number of active generations, during which the results are collected. The statistics are divided in batches, and by default each generation forms its own batch.
Notes:
- Using the pop card sets the mode to criticality source simulation. External source simulation is invoked using set nps.
- Convergence of fission source can be monitored using Shannon entropy (input parameters set his and set entr).
- See detailed description on the statistical effects of batching.
set tcut
set tcut <tmax>
Sets time cut-off for neutrons and photons. Input values:
| <tmax> | : time limit for simulated particle histories (in seconds) |
The time cut-off can be used in both neutron and photon external source simulations, to limit the length of particle histories.
Notes:
- Time or generation cut-off (set gcut) is always needed for neutron external source simulations in super-critical systems.
- Note to developers: this should take independent values for photons and neutrons
