For heavier isotopes, this approach would lead to very large files with tens of thousands of energy points. ENDF-format nuclear data evaluations combat this problem by using resonance cross section formulas to calculate the elastic, capture, and fission cross sections over a defined "resonance range." The parameters to use in these formulas are given in File 2 using MT=151. The ENDF-6 format allows the following options:

• Single-Level Breit-Wigner
• Multi-Level Breit-Wigner
• Reich-Moore
• Adler-Adler
• Hybrid R-Function
• Generalized R-Matrix

At higher energies and in heavier isotopes, the resonances get to be so closely spaced that they can no longer be resolved into separate peaks experimentally. In this "unresolved range," the ENDF format provide three ways to provide average resonance parameters to be used with statistical models from resonance theory to compute the cross sections:

• Energy-Independent Parameters
• Energy-Independent Parameters with Energy-Dependent Fission
• Energy-Dependent Parameters

The cross sections computed from the resonance formulas have to be combined with the cross sections given in File 3 to determine the evaluated cross sections. In most cases, the resonance energy range in File 3 for the resonance reactions (total, elastic, fission, capture) will either contain the value zero, or it will contain small positive and negative values to be added to the computed resonance cross sections as corrections. In a few unresolved cases, the unresolved resonance range in File 3 will contain the actual "infinitely dilute" cross section, which is to be multiplied by resonance shielding factors determined from the unresolved-resonance parameters.