SERPENT Guide

❮ Input File StructureCross Sections & Nuclear Data ❯

Materials & Compositions

Understanding Material Definitions

Material definition is fundamental to neutron transport calculations. In Serpent, materials specify the nuclear composition, density, and physical properties that determine how neutrons interact within different regions of your model.

The Material Card Structure

Every material in Serpent follows this basic syntax:

text
mat MATERIAL_NAME DENSITY [OPTIONS]
NUCLIDE1  FRACTION1
NUCLIDE2  FRACTION2
NUCLIDE3  FRACTION3
...

Component Breakdown:

  • MATERIAL_NAME: Unique identifier for the material
  • DENSITY: Mass density (negative) or atomic density (positive)
  • OPTIONS: Temperature, burnup flags, thermal scattering, etc.
  • NUCLIDE: Nuclear data identifier (e.g., 92235.09c for U-235)
  • FRACTION: Mass fraction (negative) or atomic fraction (positive)

Key Concept: Serpent uses the sign of density and fractions to distinguish between mass-based and atom-based specifications. This flexibility allows you to work with whatever data you have available.

Density and Composition Specifications

Mass Density and Mass Fractions

The most common approach uses mass density (g/cm³) with mass fractions:

text
% UO2 fuel with 3.1% enrichment
mat fuel -10.4   % Mass density: 10.4 g/cm³
92235.09c  0.031 % 3.1 at% U-235
92238.09c  0.969 % 96.9 at% U-238
8016.09c   2.0   % Stoichiometric oxygen (2 atoms per UO2)

Atomic Density and Atomic Fractions

Alternatively, use atomic density (atoms/barn-cm) with atomic ratios:

text
% Light water using atomic ratios
mat water -0.7   % Mass density: 0.7 g/cm³
1001.09c  2      % 2 hydrogen atoms per molecule
8016.09c  1      % 1 oxygen atom per molecule

Direct Atomic Densities

For precise control, specify atomic densities directly:

text
% Fuel with explicit atomic densities
mat fuel 0           % Zero density = use atomic densities below
92235.09c 1.20e-3    % atoms/(barn·cm)
92238.09c 2.15e-2    % atoms/(barn·cm)
8016.09c  4.67e-2    % atoms/(barn·cm)

Important: When using mass fractions (negative values), they don't need to sum to 1.0. Serpent automatically normalizes them. This is convenient when you have weight percentages that don't perfectly sum due to rounding.

Temperature and Physical Properties

Temperature significantly affects neutron cross sections, especially in the thermal range. Always specify realistic operating temperatures.

Temperature Specification

text
% Temperature specified in Kelvin
mat fuel -10.4 tmp 900   % Fuel at 900 K (627°C)
92235.09c -0.031
92238.09c -0.969
8016.09c  -2.0

mat clad -6.56 tmp 600   % Cladding at 600 K (327°C)
40000.09c -0.9816
50000.09c -0.0184

mat water -0.7 tmp 574   % Water at 574 K (301°C)
1001.09c  2
8016.09c  1

Thermal Scattering

For light nuclei in bound systems, thermal scattering data accounts for molecular binding effects:

text
% Water with thermal scattering
mat water -0.7 tmp 574 moder lwtr 1001
1001.09c  2    % Hydrogen
8016.09c  1    % Oxygen

% Thermal scattering data card
therm lwtr lwj3.11t

% Graphite with thermal scattering
mat graphite -1.7 tmp 900 moder grph 6000
6000.09c  1

therm grph grj3.11t

Common Thermal Scattering Data:

  • lwj3.11t - Hydrogen in light water
  • hwj3.11t - Deuterium in heavy water
  • grj3.11t - Carbon in graphite
  • bej3.11t - Beryllium metal
  • beo.10t - Beryllium oxide

Reactor Material Library

Here's a comprehensive library of common reactor materials with realistic compositions and properties:

Fuel Materials

UO₂ Fuel (Various Enrichments)

text
% Low enriched uranium (3.1 at%)
mat fuel_3p1 -10.4 tmp 900 burn 1
92235.09c  0.031   % U-235
92238.09c  0.969   % U-238
8016.09c   2.0     % Stoichiometric oxygen

% Medium enriched uranium (4.5 at%)
mat fuel_4p5 -10.4 tmp 900 burn 1
92235.09c  0.045
92238.09c  0.955
8016.09c   2.0

% Higher enriched uranium (8.0 at%)
mat fuel_8p0 -10.4 tmp 900 burn 1
92235.09c  0.080
92238.09c  0.920
8016.09c   2.0

MOX Fuel

text
% Mixed oxide fuel (Pu content: ~7 at%)
mat mox -10.8 tmp 900 burn 1
92238.09c  0.86    % U-238
94239.09c  0.05    % Pu-239
94240.09c  0.015   % Pu-240
94241.09c  0.005   % Pu-241
8016.09c   2.0     % Stoichiometric oxygen

Structural Materials

Zircaloy Alloys

text
% Zircaloy-4 (typical PWR cladding)
mat zirc4 -6.56 tmp 600
40000.09c -0.9816  % Zirconium
50000.09c -0.0150  % Tin
26000.09c -0.0021  % Iron
24000.09c -0.0010  % Chromium
8016.09c  -0.0012  % Oxygen

% Zircaloy-2 (typical BWR cladding)
mat zirc2 -6.56 tmp 600
40000.09c -0.9845
50000.09c -0.0150
26000.09c -0.0012
24000.09c -0.0010
28000.09c -0.0005  % Nickel

Stainless Steel

text
% 304 Stainless Steel
mat ss304 -8.0 tmp 600
26000.09c -0.695   % Iron
24000.09c -0.190   % Chromium
28000.09c -0.095   % Nickel
25055.09c -0.020   % Manganese

% 316 Stainless Steel (higher corrosion resistance)
mat ss316 -8.0 tmp 600
26000.09c -0.650
24000.09c -0.170
28000.09c -0.120
42000.09c -0.025   % Molybdenum
25055.09c -0.020
14000.09c -0.010   % Silicon
6000.09c  -0.003   % Carbon

Moderator and Coolant

Water at Different Conditions

text
% Cold water (room temperature)
mat water_cold -1.0 tmp 293 moder lwtr 1001
1001.09c  2
8016.09c  1

% PWR operating conditions
mat water_pwr -0.7 tmp 574 moder lwtr 1001
1001.09c  2
8016.09c  1

% BWR steam (simplified as low-density water)
mat steam -0.1 tmp 550 moder lwtr 1001
1001.09c  2
8016.09c  1

therm lwtr lwj3.11t

Borated Water

text
% Borated water (1000 ppm natural boron)
mat bwater -0.7 tmp 574 moder lwtr 1001
1001.09c  2.0
8016.09c  1.0
5010.09c  0.000199  % B-10 (19.9% of natural boron)
5011.09c  0.000801  % B-11 (80.1% of natural boron)

% Borated water (2000 ppm)
mat bwater_2000 -0.7 tmp 574 moder lwtr 1001
1001.09c  2.0
8016.09c  1.0
5010.09c  0.000398
5011.09c  0.001602

therm lwtr lwj3.11t

Control Materials

text
% Ag-In-Cd control rod material
mat aic -10.17 tmp 600
47000.09c -0.80    % Silver
49000.09c -0.15    % Indium
48000.09c -0.05    % Cadmium

% Boron carbide (B4C)
mat b4c -2.52 tmp 600
5010.09c  3.128    % B-10 (natural abundance)
5011.09c  12.598   % B-11
6000.09c  4.0      % Carbon

% Hafnium control rod
mat hafnium -13.3 tmp 600
72000.09c -1.0

Advanced Material Features

Burnup and Depletion

For fuel cycle studies, mark fissile materials for burnup tracking:

text
% Fuel material with burnup tracking
mat fuel -10.4 tmp 900 burn 1
92235.09c -0.031
92238.09c -0.969
8016.09c  -2.0

% Multiple burnup zones in same material
mat fuel_zone1 -10.4 tmp 900 burn 1
92235.09c -0.031
92238.09c -0.969
8016.09c  -2.0

mat fuel_zone2 -10.4 tmp 900 burn 2  % Different burn zone
92235.09c -0.031
92238.09c -0.969
8016.09c  -2.0

Material Mixtures

Create composite materials by mixing pre-defined materials:

text
% Define base materials
mat uo2_high -10.4 tmp 900
92235.09c -0.045
92238.09c -0.955
8016.09c  -2.0

mat uo2_low -10.4 tmp 900
92235.09c -0.030
92238.09c -0.970
8016.09c  -2.0

% Create mixed fuel
mat fuel_mixed -10.4 tmp 900 burn 1
mix uo2_high 0.3   % 30% high enrichment
mix uo2_low  0.7   % 70% low enrichment

Volume and Mass Specification

For reaction rate calculations and inventory tracking:

text
% Material with explicit volume
mat fuel -10.4 tmp 900 vol 125.6 burn 1  % Volume in cm³
92235.09c -0.031
92238.09c -0.969
8016.09c  -2.0

% Material with explicit mass
mat fuel -10.4 tmp 900 mass 1306.2 burn 1  % Mass in grams
92235.09c -0.031
92238.09c -0.969
8016.09c  -2.0

Common Mistakes and Best Practices

❌ Common Mistakes

  • Temperature units: Always use Kelvin, not Celsius
  • Density signs: Mixing positive and negative density conventions
  • Missing thermal scattering: Forgetting thermal data for light nuclei
  • Unrealistic densities: Using handbook values without temperature correction
  • Isotope notation: Using wrong nuclear data identifiers

✅ Best Practices

  • Organize materials: Group by type (fuel, structure, coolant)
  • Use comments: Document the source of composition data
  • Validate densities: Check that final densities are reasonable
  • Temperature consistency: Use realistic operating temperatures
  • Library management: Maintain a validated material database

Material Validation

Always verify your material definitions before using them in production runs:

text
% Add a detector to check material properties
det material_check dm fuel de myenergy

% Energy grid for spectrum analysis
ene myenergy 3 500 1e-9 20.0

% This will show if your material behaves as expected

Pro Tip: Create a simple infinite medium test case for each new material to verify its neutron multiplication properties and cross-section behavior before using it in complex geometries.

Related Topics

❮ Input File StructureCross Sections & Nuclear Data ❯