Cross Section Data in SCONE

Nuclear cross sections form the foundation of Monte Carlo neutron transport simulations. These fundamental quantities describe how neutrons interact with different atomic nuclei, determining their behavior as they travel through materials. SCONE uses this data to calculate mean free paths, determine interaction types, and generate secondary particles with appropriate properties.

When a neutron travels through matter, cross section data guides every aspect of its journey. The data determines how far the neutron will travel before interacting, what type of interaction occurs (such as scattering or absorption), and the resulting changes in neutron energy and direction.

SCONE reads cross-section data from the nuclearData block. The handles section declares database connections. For continuous-energy neutron transport, use an ACE-format library via aceNeutronDatabase. Point aceLibrary to your ACE cross-section file (e.g., from JEF, ENDF/B, or JENDL processed with NJOY).

nuclearData {
  handles {
    ce { type aceNeutronDatabase; aceLibrary /path/to/JEF311.aceXS; }
  }
}

SCONE primarily works with ACE (A Compact ENDF) format data, optimized for Monte Carlo calculations. This format efficiently organizes nuclear data into a structured form, including energy grids, reaction cross sections, and distribution data for secondary particles.

// ACE format description (conceptual)
// Header: ZAID.NN  nuclide name  library  temperature
// 92235.06  U-235 ENDF/B-VII.1 293.6K
//
// ACE files contain: energy grids, cross sections (total, elastic, fission, etc.),
// angular distributions, energy distributions, and probability tables for URR.
// Process ENDF data to ACE using NJOY or similar codes.

Processing tools play a crucial role in preparing nuclear data for SCONE. The industry-standard NJOY code converts ENDF data to ACE format, handling Doppler broadening and thermal scattering law processing. SCONE expects ACE libraries in the format produced by such processing.

Nuclides are identified by ZAID (Z-A ID) with a temperature index. The format is ZZZAAA.NN: ZZZ is the atomic number (padded to 3 digits), AAA is the mass number (padded to 3 digits), and NN is the temperature index. For example, 92235.06 denotes U-235 at 600K (index 06). The temperature index corresponds to pre-processed data at specific temperatures in the ACE library.

// ZAID.NN format: ZZZAAA.NN
// ZZZ = element (Z), AAA = mass number, NN = temperature index
// Example: 92235.06 = U-235 at 600K (index 06)
// Example: 1001.06  = H-1 at 600K
// Example: 8016.06  = O-16 at 600K

// In material composition:
Water {
  temp 600;
  composition {
    5010.06 7.9714E-06;   // B-10
    5011.06 3.2247E-05;   // B-11
    1002.06 7.7035E-06;   // H-2
    8016.06 2.4673E-02;   // O-16
  }
}

Material compositions are defined in the nuclearData.materials block. Each material specifies a temperature and a composition of ZAID.NN pairs with number densities. The ZAID values must match nuclides available in the ACE library referenced in handles.

nuclearData {
  handles {
    ce { type aceNeutronDatabase; aceLibrary /path/to/your/library.aceXS; }
  }
  materials {
    UO2-31 {
      temp 600;
      composition {
        8016.06  4.5853E-02;
        92235.06 7.2175E-04;
        92238.06 2.2253E-02;
      }
    }
  }
}

Temperature significantly influences neutron cross sections, particularly in the resonance region. SCONE uses pre-processed ACE data at discrete temperatures. The ZAID temperature index (e.g., .06 for 600K) selects the appropriate data. Ensure your ACE library was processed at temperatures that bracket your simulation conditions.

Maintain consistency across data libraries. Use ZAID indices that match your material temperatures. Verify that all nuclides in your material compositions exist in the ACE library. Document the library path and version for reproducibility.