SERPENT Guide

❮ Cross Sections & Nuclear DataSimple Geometries ❯

Source Definition

Understanding Neutron Sources

Neutron sources define where and how neutrons are introduced into your simulation. The source specification depends on your problem type: criticality calculations use an initial fission source that evolves during the simulation, while fixed source problems require explicit spatial and energy distributions.

Key Concept: Serpent automatically handles many source details for you. For criticality calculations, you often only need to specify the calculation mode, while for shielding and activation problems, you define detailed source characteristics.

Calculation Modes

Serpent supports different calculation modes that determine how neutron sources are treated:

Available Calculation Modes:

  • Criticality (kcode): Eigenvalue calculations for reactor physics
  • Fixed source: External neutron source calculations
  • External source: Time-dependent or coupled calculations
  • Burnup: Depletion calculations with fission source evolution

Criticality Calculations

For reactor physics calculations, Serpent typically runs in criticality mode where the neutron source comes from fission reactions. The initial source distribution is iterated until convergence to the fundamental eigenmode.

Basic Criticality Setup

text
% Basic criticality calculation
set pop 10000 500 100  % Neutrons per cycle, active cycles, inactive cycles

% Optional: provide an initial source guess from a previous run
set srcfile "initial_source.srv"

Best Practice: Use 100-200 inactive cycles for source convergence in complex geometries. Monitor the Shannon entropy and source distribution plots to verify convergence before collecting statistics.

Source Convergence Monitoring

text
% Enable source convergence diagnostics
set entropy 1          % Calculate Shannon entropy
set entropymat fuel    % Monitor entropy in fuel materials

% Source point output for visualization
set srcfile "source_dist" 1  % Write source distribution after run

Fixed Source Calculations

Fixed source calculations are used for shielding analysis, activation studies, and external source problems. You must explicitly define the neutron source characteristics including spatial distribution, energy spectrum, and angular distribution.

Basic Source Definition

text
% Set calculation mode to fixed source
set nps 1000000        % Number of source neutrons to simulate

% Basic point source
src source1
  sp  0.0 0.0 0.0      % Position (x, y, z) in cm
  se  1.0              % Monoenergetic 1 MeV neutrons
  sd  1.0              % Isotropic direction
end

% Alternative: directional source
src source2  
  sp  0.0 0.0 0.0      % Source position
  se  14.1             % 14.1 MeV (D-T fusion neutrons)
  su  0.0 0.0 1.0      % Direction vector (toward +z)
  si  -1.0 1.0         % Cosine distribution limits
  sp  1.0              % Uniform cosine distribution  
end

Volumetric Sources

text
% Uniform volume source in a cylinder
src volume_source
  sx  -10.0 10.0       % X limits (cm)
  sy  -10.0 10.0       % Y limits (cm) 
  sz  0.0 20.0         % Z limits (cm)
  sc  1                % Cylindrical source
  sr  0.0 5.0          % Radius limits (cm)
  se  2.45             % Cf-252 average energy (MeV)
  sd  1.0              % Isotropic
end

% Material-based source (fuel assemblies)
src fuel_source
  sm  fuel             % Source in 'fuel' material
  se  1.0              % 1 MeV neutrons
  sd  1.0              % Isotropic
  sp  0.0 0.0 0.0      % Reference position (for relative sampling)
end

Note: Material-based sources automatically distribute neutrons proportionally to the material volume throughout the geometry. This is ideal for modeling distributed activation or fission sources.

Energy Distributions

Realistic energy distributions are crucial for accurate source calculations. Serpent supports various energy spectrum formats including tabulated distributions, analytical functions, and standard reactor physics spectra.

Common Energy Spectra

text
% Watt fission spectrum (typical for thermal reactors)
src fission_spectrum
  sp  0.0 0.0 0.0      % Source position
  se  -3               % Watt spectrum (negative value)
  sw  0.988 2.249      % Watt parameters (a=0.988, b=2.249 for U-235)
  sd  1.0              % Isotropic
end

% Maxwellian thermal spectrum
src thermal_source
  sp  0.0 0.0 0.0      % Source position  
  se  -2               % Maxwellian spectrum
  sw  0.0253           % Temperature in eV (300K thermal)
  sd  1.0              % Isotropic
end

% Tabulated energy distribution
src tabulated_source
  sp  0.0 0.0 0.0      % Source position
  se  -1               % Tabulated spectrum
  sf  "energy_dist.dat" % File with energy-probability pairs
  sd  1.0              % Isotropic  
end

Energy Distribution File Format

text
% Energy distribution file format (energy_dist.dat)
% Energy (MeV)  Probability
1.0E-8          0.001
1.0E-6          0.01  
1.0E-4          0.05
0.01            0.10
0.1             0.15
1.0             0.20
5.0             0.25
10.0            0.15
14.1            0.089  % D-T peak
20.0            0.001

Performance Tip: Use energy cutoffs to avoid tracking low-energy neutrons that don't contribute to your results. Set appropriate lower energy limits with theset ncut command.

Practical Source Examples

Reactor Startup Source

text
% Antimony-beryllium startup source in reactor core
src startup_source
  sx  -15.0 15.0       % Core x-limits (cm)
  sy  -15.0 15.0       % Core y-limits (cm)
  sz  0.0 365.0        % Active core height (cm)
  sm  fuel             % Source in fuel material only
  se  1.0              % 1 MeV average (Sb-Be source)
  sd  1.0              % Isotropic
  si  fuel_importance.dat  % Importance weighting file (optional)
end

% Source normalization
set power 3400         % 3400 MW thermal power
set powdens 1 fuel     % Power produced in fuel material

Spent Fuel Cask Source

text
% Spent fuel neutron source for shielding analysis
src spent_fuel
  sx  -20.0 20.0       % Fuel assembly x-limits
  sy  -20.0 20.0       % Fuel assembly y-limits  
  sz  0.0 400.0        % Fuel stack height
  sm  spent_fuel       % Source in spent fuel material
  se  -3               % Watt spectrum
  sw  1.025 2.926      % Watt parameters for Pu-239 
  sd  1.0              % Isotropic
  sp  gamma            % Include gamma sources
  sg  "decay_gamma.dat" % Decay gamma spectrum
end

% Source strength (neutrons/sec)
set nps 1.5E8          % 1.5×10⁸ neutrons total
set srcrate 1.2E6      % 1.2×10⁶ n/sec source rate

Calibration Source

text
% Cf-252 calibration source for detector response
src cf252_source
  sp  0.0 0.0 5.0      % 5 cm above detector
  se  -3               % Watt spectrum
  sw  1.025 2.926      % Cf-252 Watt parameters
  sd  1.0              % Isotropic
  si  4π 1.0           % Full solid angle
  sp  gamma            % Include spontaneous fission gammas
  sg  -4               % Prompt fission gamma spectrum
end

% Point source normalization  
set srcrate 2.3E6      % 2.3×10⁶ fissions/sec (1 μg Cf-252)

Advanced Source Features

Time-Dependent Sources

text
% Pulsed neutron source
src pulsed_source
  sp  0.0 0.0 0.0      % Source position
  se  14.1             % 14 MeV D-T neutrons
  sd  1.0              % Isotropic
  st  -1               % Time-dependent
  stf "pulse_profile.dat"  % Time distribution file
end

% Time distribution file format:
% Time (s)    Relative Intensity
0.0           1.0
1.0E-6        1.0      % 1 μs pulse width
1.0E-5        0.0      % Pulse ends
1.0E-3        0.0      % Measurement period

Multiple Source Definition

text
% Multiple independent sources
src control_rod_source
  sp  5.0 5.0 100.0    % Control rod position
  se  1.0              % 1 MeV
  sd  1.0              % Isotropic
  sw  0.1              % 10% of total source strength
end

src detector_calibration  
  sp  -10.0 -10.0 50.0 % Calibration position
  se  0.0253           % Thermal neutrons
  sd  1.0              % Isotropic
  sw  0.9              % 90% of total source strength
end

% Total source normalization
set nps 1000000        % Total neutrons from all sources

Advanced Feature: Multiple sources allow you to model complex scenarios like reactor startup with both intrinsic and external sources, or calibration setups with multiple reference sources.

Related Topics

❮ Cross Sections & Nuclear DataSimple Geometries ❯