Source Definition

Neutron Sources

Neutron sources define where and how neutrons enter a Monte Carlo simulation. In criticality calculations, the fission source distribution evolves iteratively until it converges to the fundamental eigenmode — the analyst typically needs only population parameters. In fixed-source problems (shielding, activation, detector response), the spatial distribution, energy spectrum, and angular distribution must be specified explicitly.

Serpent supports criticality (eigenvalue) mode, fixed-source mode for external source problems, external source mode for time-dependent or coupled calculations, and burnup mode where the fission source evolves with changing material compositions.

Criticality Calculations

In criticality mode, Serpent begins with an initial fission source guess and iterates through inactive cycles to converge the source distribution before collecting statistics during active cycles. The set pop card controls neutrons per cycle, active cycles, and inactive cycles. An initial source from a previous calculation can accelerate convergence for complex geometries.

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

% Serpent can read an initial source distribution from a previous
% run using the set srcpoint card (see Serpent Wiki for details).

Shannon entropy of the fission source provides a quantitative convergence measure: once it stabilizes around a constant value, the source has reached a stationary distribution. Serpent calculates Shannon entropy automatically and can restrict the calculation to specific materials (such as fuel) for more meaningful diagnostics.

% Enable source convergence diagnostics
set entr 20 20 1       % Shannon entropy mesh (20x20x1 spatial bins)

% Serpent writes source point files automatically at the end of
% each run, which can be used to restart subsequent calculations.

Fixed-Source Calculations

Fixed-source calculations apply to shielding, activation, detector response, and any problem with an externally defined neutron source. The analyst must specify the source's spatial location, energy spectrum, angular distribution, and intensity. The src card supports point, volumetric, surface, and material-distributed sources.

A point source is the simplest source definition, placing all source neutrons at a single location in space. The sp sub-card specifies the position, se defines the energy, and sd controls the angular distribution. Emission is isotropic by default when sd is omitted. The sd sub-card specifies a direction vector (sd UX UY UZ) for anisotropic or beam-like sources.

% 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

% Alternative: directional source
src source2  
  sp  0.0 0.0 0.0      % Source position
  se  14.1             % 14.1 MeV (D-T fusion neutrons)
  sd  0.0 0.0 1.0      % Direction vector (toward +z)

Volumetric sources distribute neutrons throughout a spatial region — appropriate for activation products, spontaneous fission in spent fuel, or startup fission sources. The region can be defined by coordinate limits, cylindrical boundaries, or a material name. Material-based sources distribute neutrons proportionally to the material volume throughout the geometry, automatically following the spatial distribution of that material.

% Volume source (simplified)
src volume_source
  sp  0.0 0.0 10.0     % Center position (cm)
  se  2.13             % Cf-252 average energy (MeV)
  % For volumetric spatial distributions (box, cylinder, etc.),
  % consult the Serpent Wiki for the full 'src' card syntax.

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

Energy Distributions

The source energy spectrum directly affects the transport solution and all downstream tallies. Serpent supports monoenergetic sources, analytical distributions (Watt fission spectrum, Maxwellian thermal spectrum), and tabulated energy-probability distributions.

The Watt fission spectrum is the standard choice for modeling fission neutron sources from thermal reactor fuels and is selected with se -3, which uses built-in parameters for U-235 thermal fission. The Maxwellian spectrum, selected with se -2, is appropriate for thermalized neutron sources. Tabulated distributions provide maximum flexibility, allowing the analyst to specify an arbitrary energy-probability histogram.

% Watt fission spectrum (typical for thermal reactors)
src fission_spectrum
  sp  0.0 0.0 0.0      % Source position
  se  -3               % Watt fission spectrum (built-in U-235 defaults)

% Maxwellian thermal spectrum
src thermal_source
  sp  0.0 0.0 0.0      % Source position  
  se  -2               % Maxwellian spectrum

% Tabulated energy distribution
src tabulated_source
  sp  0.0 0.0 0.0      % Source position
  se  -1               % Tabulated spectrum
  % Consult the Serpent Wiki for specifying tabulated energy distributions.

For tabulated energy distributions, consult the Serpent Wiki for the correct syntax and file format. Generally, tabulated distributions use energy-probability pairs that Serpent normalizes automatically.

% 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

Practical Source Examples

A reactor startup source models the initial neutron population during commissioning, where an antimony-beryllium or californium source provides the seed neutrons. The source is distributed throughout the fuel material and normalized to the reactor thermal power.

% Antimony-beryllium startup source in reactor core
src startup_source
  sm  fuel             % Source in fuel material only
  sp  0.0 0.0 182.5    % Core center position (cm)
  se  0.024            % 0.024 MeV average (Sb-Be source)

% Source normalization
set power 3.4E9        % 3400 MW thermal power (in watts)

Spent fuel cask shielding analysis requires modeling both neutron and gamma sources from the decay of fission products and actinides. The neutron source typically follows a Watt spectrum with parameters appropriate for spontaneous fission of plutonium isotopes, while the gamma source is specified through a decay gamma spectrum file derived from the fuel's burnup history.

% Spent fuel neutron source for shielding analysis
src spent_fuel
  sm  spent_fuel       % Source in spent fuel material
  sp  0.0 0.0 200.0    % Center position (cm)
  se  -3               % Watt fission spectrum
  % For gamma sources, consult the Serpent Wiki for the
  % full 'src' card syntax and photon transport options.

set nps 1000000        % Total neutrons to simulate

Detector calibration simulations model a known source, such as Cf-252, positioned near a detector to calculate the detector's response function. The Cf-252 spontaneous fission source emits neutrons with a well-characterized Watt spectrum and also produces prompt fission gammas that may contribute to the detector signal.

% Cf-252 calibration source for detector response
src cf252_source
  sp  0.0 0.0 5.0      % 5 cm above detector
  se  -3               % Watt fission spectrum (Cf-252)

set nps 1000000        % Total neutrons to simulate

Advanced Source Features

Serpent's dynamic simulation modes can model pulsed neutron experiments, accelerator-driven systems, and other time-dependent scenarios. Consult the Serpent Wiki for available time-dependent source options.

% D-T fusion neutron source
src dt_source
  sp  0.0 0.0 0.0      % Source position
  se  14.1             % 14.1 MeV D-T neutrons

% For time-dependent or pulsed source simulations, consult
% the Serpent Wiki for available dynamic simulation modes.

Multiple independent sources can be defined in a single simulation, each with its own spatial, energy, and angular distributions. Weight factors control the relative source strengths, and Serpent samples from each source according to these weights.

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

src detector_calibration  
  sp  -10.0 -10.0 50.0 % Calibration position
  se  2.53E-8          % Thermal neutrons (0.0253 eV)
  sw  0.9              % 90% of total source strength

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