MCNP Guide
Materials in MCNP
Defining the physical composition of your model
Understanding Material Definitions
Material cards tell MCNP what atoms are present in each region of your model and in what proportions. Each material gets a unique number that you reference in your cell definitions. The key to effective material modeling is understanding how to specify isotopic compositions and their associated cross-section data.
Basic Material Format
c Material card structure
m1 92235.80c 0.045 $ U-235 (4.5 at% enriched)
92238.80c 0.955 $ U-238 (remainder)
8016.80c 2.000 $ O-16 (stoichiometric oxygen)
c Breaking down the format:
c m1 = material number 1
c 92235.80c = uranium-235 with 80-series cross sections
c 0.045 = atom fraction (positive number)
c $ comment explaining the entryThe material number (m1) connects to your cell definitions. The isotope identifier (92235.80c) specifies both the nuclear species and the cross-section library. The fraction tells MCNP how much of each isotope is present. Positive numbers indicate atom fractions (or atom ratios), while negative numbers indicate weight fractions.
Cross-Section Libraries
The cross-section library identifier (like .80c) determines which nuclear data MCNP uses for each isotope. Different libraries represent different evaluations and temperature ranges. Choosing the right library affects the accuracy of your results.
c Common cross-section libraries
m1 92235.80c 0.045 $ ENDF/B-VIII.0 at 293.6K
92238.80c 0.955 $ Standard temperature
m2 92235.81c 0.045 $ ENDF/B-VIII.0 at 600K
92238.81c 0.955 $ Elevated temperature
m3 92235.70c 0.045 $ ENDF/B-VII.0 evaluation
92238.70c 0.955 $ Room temperatureThe .80c suffix represents room temperature (293.6K) data from ENDF/B-VIII.0, while .81c provides the same evaluation at 600K. The temperature of your cross-section library should match the operating temperature of your material. For most reactor applications, you'll use .80c for room temperature components and .81c for heated materials like fuel. Note: exact suffix-to-library mappings depend on your installation's xsdir file.
Practical Material Examples
Nuclear Fuel
Nuclear fuel requires precise isotopic specifications because small changes in enrichment significantly affect reactivity. Here's how to define typical reactor fuel.
c LEU UO2 fuel (4.5% enriched, 10.4 g/cm³)
m1 92235.80c 0.045 $ U-235 enrichment
92238.80c 0.955 $ U-238 remainder
8016.80c 2.000 $ Oxygen (stoichiometric UO2)
c Fresh fuel vs. depleted fuel
m2 92235.80c 0.031 $ Depleted U-235 (3.1%)
92238.80c 0.969 $ Depleted U-238
8016.80c 2.000 $ Oxygen
94239.80c 0.0045 $ Pu-239 from burnup
94240.80c 0.0015 $ Pu-240 from burnupFresh fuel contains only uranium and oxygen in the UO₂ structure. The uranium fractions represent the enrichment level, while oxygen maintains the stoichiometric ratio. Depleted fuel includes plutonium isotopes created during reactor operation, which significantly affect the neutron balance.
Moderator and Coolant
Water serves as both moderator and coolant in most power reactors. The presence of boron for reactivity control and the temperature-dependent density require careful specification.
c Borated water (1200 ppm boron, 580K)
m3 1001.80c 2.0 $ Hydrogen in H2O
8016.80c 1.0 $ Oxygen in H2O
5010.80c 0.00040 $ B-10 (20% of natural boron)
5011.80c 0.00160 $ B-11 (80% of natural boron)
mt3 lwtr.20t $ Thermal scattering treatment
c Heavy water moderator (99.75% D2O)
m4 1002.80c 1.9975 $ Deuterium (99.75%)
1001.80c 0.0025 $ Hydrogen (0.25% light water)
8016.80c 1.0 $ Oxygen
mt4 hwtr.20t $ Heavy water thermal treatmentThe boron concentration in pressurized water reactors typically ranges from 0 to 2000 ppm, with natural boron isotopic ratios. The mt3 card adds thermal neutron scattering data, which is crucial for accurate neutron thermalization modeling. Heavy water reactors require deuterium specification and appropriate thermal treatment data.
Structural Materials
Structural materials like cladding and vessel steel affect neutron absorption and must be modeled accurately. These materials often contain multiple alloying elements that influence neutron behavior.
c Zircaloy-4 cladding (6.55 g/cm³)
m5 40090.80c 0.5145 $ Zr-90 (natural zirconium)
40091.80c 0.1122 $ Zr-91
40092.80c 0.1715 $ Zr-92
40094.80c 0.1738 $ Zr-94
40096.80c 0.0280 $ Zr-96
50120.80c 0.0100 $ Tin (corrosion resistance)
c Stainless steel 316 (8.0 g/cm³)
m6 26056.80c 0.6800 $ Iron (68%)
24052.80c 0.1700 $ Chromium (17%)
28058.80c 0.1200 $ Nickel (12%)
42095.80c 0.0250 $ Molybdenum (2.5%)
14028.80c 0.0050 $ Silicon (0.5%)Zircaloy-4 cladding requires all zirconium isotopes because their different absorption cross-sections affect neutron economy. The tin addition improves corrosion resistance but also absorbs neutrons. Stainless steel components need complete alloy specifications because chromium and nickel have significant absorption cross-sections that affect neutron flux distributions.
Temperature and Density Effects
Material properties change with temperature and pressure, affecting both cross-sections and densities. MCNP provides several ways to handle these variations accurately.
c Hot fuel (1200K operating temperature)
m10 92235.81c 0.045 $ U-235 with 600K cross sections
92238.81c 0.955 $ U-238 with 600K cross sections
8016.81c 2.000 $ O-16 with 600K cross sections
TMP 1.0341e-7 $ Doppler broadening at 1200K
c Temperature-dependent water density
c (density changes from 1.0 g/cm³ at 293K to 0.73 g/cm³ at 580K)
m11 1001.80c 2.0 $ Hydrogen
8016.80c 1.0 $ Oxygen
mt11 lwtr.20t $ Thermal treatment at 580K
c Helium gap (density set on cell card)
m12 2004.80c 1.0 $ Helium-4
c Density on cell card reflects fill-gas pressure
c e.g. 5 12 -0.0025 ... $ He at ~15 atmHigh-temperature fuel requires cross-section libraries that match the operating temperature, plus TMP cards for Doppler broadening effects. The density changes in your cell definitions must reflect thermal expansion. For pressurized gases like helium in fuel rod gaps, specify the appropriate mass density directly on the cell card to reflect fill-gas pressure and temperature.
Common Mistakes and Solutions
Material definition errors can cause significant problems in MCNP calculations. The most common mistake is inconsistent normalization of fractions. When using weight fractions (negative numbers), ensure they sum to 1.0 for each material. When using atom fractions (positive numbers), verify the ratios match your intended composition.
Cross-section library mismatches create subtle errors. Using room temperature libraries for hot materials underestimates Doppler broadening effects. Missing thermal scattering treatments for moderator materials can significantly affect neutron thermalization and multiplication factors.
Always verify your material densities match the values specified in your cell definitions. The material composition defines what atoms are present, but the cell definition determines how many atoms per unit volume. Inconsistent densities between material specifications and cell definitions lead to incorrect neutron interaction rates.