MCNP Guide
Cross Sections in MCNP
Managing nuclear data for accurate simulations
Nuclear Data Libraries
Cross-section libraries contain evaluated nuclear data that determines how particles interact with materials. Choosing the right library affects both accuracy and computational efficiency.
Library Identification
c Common nuclear data libraries
m1 92235.80c 1.0 $ U-235 from ENDF/B-VIII.0 (293.6K)
m2 92235.70c 1.0 $ U-235 from ENDF/B-VII.0 (293.6K)
m3 92235.71c 1.0 $ U-235 from ENDF/B-VII.1 (293.6K)
c ZAID format: ZZZAAA.XXc
c ZZZ = atomic number (92 for uranium)
c AAA = mass number (235)
c XX = library version (installation-dependent)
c c = continuous energyUse consistent libraries throughout your model. ENDF/B-VIII.0 (.80c) is the latest US standard, while ENDF/B-VII.1 (.70c) has extensive validation.
Temperature Treatment
Nuclear cross sections depend on material temperature due to Doppler broadening. MCNP offers several approaches to handle temperature effects accurately.
Pre-broadened Libraries
c Fuel at operating temperature (600K)
m1 92235.81c -0.04 $ U-235 at 600K
92238.81c -0.96 $ U-238 at 600K
8016.81c -0.135 $ O-16 at 600K
c Room temperature components (293.6K)
m2 40090.80c -0.98 $ Zr-90 cladding
40091.80c -0.02 $ Zr-91 cladding
c Available pre-broadened temperatures:
c .80c = 293.6K (room temperature)
c .81c = 600K (typical fuel temperature)
c .82c = 900K (high temperature fuel)
c .84c = 1200K (accident conditions)Pre-broadened data provides the most accurate temperature treatment. Use .81c for typical reactor fuel temperatures.
Runtime Temperature Adjustment
c Specify cell temperatures on cell cards (kT in MeV)
c Cell 1: tmp=6.90e-8 (800K)
c Cell 2: tmp=5.17e-8 (600K)
c Cell 3: tmp=2.53e-8 (293.6K)
c Or as a single data card listing one value per cell:
TMP 6.90e-8 5.17e-8 2.53e-8
c Enable Doppler-broadening rejection correction
DBRC 92238 92235 $ Apply DBRC to these nuclides
c Temperature conversion: kT(MeV) = 8.617e-11 × T(K)Use TMP cards when pre-broadened data isn't available. DBRC improves accuracy for resonance absorption in the thermal and epithermal range.
Thermal Scattering Data
Thermal neutrons in bound materials require special scattering treatment. S(α,β) data accounts for molecular binding and lattice effects.
c Water moderator with thermal scattering
m1 1001.80c 2 $ Hydrogen in H2O
8016.80c 1 $ Oxygen in H2O
mt1 lwtr.10t $ Light water S(α,β) at 293K
c Graphite moderator
m2 6000.80c 1 $ Carbon
mt2 grph.10t $ Graphite S(α,β)
c Heavy water
m3 1002.80c 2 $ Deuterium in D2O
8016.80c 1 $ Oxygen in D2O
mt3 hwtr.10t $ Heavy water S(α,β)
c Polyethylene shielding
m4 1001.80c 4 $ Hydrogen in (CH2)n
6000.80c 2 $ Carbon in (CH2)n
mt4 poly.10t $ Polyethylene S(α,β)
c Temperature-dependent alternatives (use ONE mt per material):
c mt1 lwtr.01t $ Light water at 350K (pick one)
c mt1 lwtr.02t $ Light water at 400K (pick one)Always use S(α,β) data for moderator materials. The .10t suffix indicates room temperature thermal data. Match temperatures to your operating conditions.
Advanced Cross Section Features
Specialized treatments improve accuracy for specific energy ranges and applications. Use these features when standard cross sections aren't sufficient.
Unresolved Resonance Treatment
c Probability tables for unresolved resonance region
c Controlled via the PHYS:n card (3rd entry, iunr):
PHYS:n 20 0 -1 $ iunr=-1 enables probability
c tables for all nuclides
c
c Essential for fast-spectrum and intermediate-energy
c reactor calculations where unresolved resonances
c affect self-shielding.Setting iunr = −1 on the PHYS:n card enables probability table sampling in the unresolved resonance region. Essential for accurate fast reactor calculations.
Photon Production
c Control photon production from neutron reactions
LCA 2 2 2 2 2 $ Photon production ON for materials 1-5
LCB 0 0 0 0 0 $ Neutron production OFF for materials 1-5
c Detailed photonuclear physics
PIKMT 100 102 103 $ Track specific photonuclear reactions
c | | |
c | | +-- (γ,n) reaction
c | +-- (γ,2n) reaction
c +-- Total photonuclear
c Energy thresholds
PHYS:p 20 0 1 1 $ Enable photonuclear to 20 MeVControl secondary particle production for coupled neutron-photon calculations. Essential for shielding and activation studies.
Library Selection Guidelines
Reactor Physics
Primary: ENDF/B-VII.1 (.70c)
Alternative: ENDF/B-VIII.0 (.80c)
Thermal: S(α,β) required
Temperature: Pre-broadened preferred
Shielding Analysis
Primary: ENDF/B-VIII.0 (.80c)
Photons: Coupled transport
Thermal: S(α,β) for moderators
Temperature: Room temperature OK
Fast Systems
Primary: ENDF/B-VIII.0 (.80c)
Special: URRG for actinides
Thermal: Not critical
Temperature: High-T data needed
Dosimetry
Primary: ENDF/B-VIII.0 (.80c)
Photons: Detailed physics
Thermal: S(α,β) if present
Temperature: Match conditions
Common Issues and Solutions
Troubleshooting Cross Section Problems
Missing isotope data
Use natural element data (e.g., 6000.80c for carbon) when specific isotopes aren't available.
Temperature mismatch
Ensure S(α,β) data temperature matches material temperature. Use closest available data.
Library inconsistencies
Use the same library version for all isotopes in a material to avoid evaluation inconsistencies.
Energy range issues
Verify cross section energy limits match your problem's energy range. Extend if necessary.
Best Practices
Maintain consistency in library selection across your model. Document your cross section choices and validate results against benchmarks when possible.
For production calculations, use well-validated libraries like ENDF/B-VII.1. Newer isn't always better - some applications benefit from older, extensively tested evaluations.