Evaluated Nuclear Data File (ENDF)

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This page was designed to help a user without any knowledge of ENDF format. A user with some knowledge of ENDF format may benefit from Extended Retrieval and from Advanced Retrieval. Star (*) denotes a wild card in input parameters.


Target  |  Reaction |  Quantity  |  Library  
  Sub-library  |  MT number  |  MF number  |  Product  |  Laboratory  |  Author  
  Examples  |  How to?  |  What is?  |  Why?

Target
In the ENDF or SIGMA interface the word 'target' is used in the usuall nuclear physics meaning, i.e., it specifies a nuclide with which an incident paticle (projectile) is interacting. We choose 'target' instead of the ENDF-6 term 'material' since we believe the former is more familiar and informative to the users with nuclear physics background. We admit, however, that for certain sublibraries, such as decay sublibrary or spontanous fission yields sublibrary, the term 'material' is more appropriate since target implies existence of a projectile, which is not involved in these specific cases.

Reaction
List of reactions is closely related to MT numbers. User typically requests n,*; n,tot; n,g; n,f; n,el; n,inl; n,2n; n,a; n,nu*; p,*;�, where n = neutron, * = anything, tot= total, g = gamma, f = fission, el = elastic, inl = inelastic, a = alpha, nu* = nu-bar with all options (prompt, total, delayed), p = proton.

Quantity
List of quantities is defined by MF numbers. User typically requests sig; da; de; res; cov*, where sig = reaction cross-section, de = energy spectrum of emitted particles, da = angular distribution of emitted particles, res = neutron resonance parameters, cov* = covariances with all options (cross sections, nu-bar, neutron resonance parameters).

Library
Libraries available via the retrieval interface are the major national projects and a few most important special purpose libraries. All these libraries are in the current ENDF-6 format, which elimintaes older versions such as ENDF/B-I, -II, -III, -IV and -V. Most recent versions of the following libraries are included:

Sub-library (Projectile)
ENDF is organized into sub-libraries that usually represent projectiles such as n, p and g, where n=neutron, p=proton, g=gamma. There are other sub-libraries such as decay=decay data, fission yields, ... The ENDF/B-VII.1 and -VII.0 libraries contain the most extended variety of sublibraries:

MT number (Ejectile)
MT numbers provide the most detailed definition of emitted particles (ejectiles): 1=total, 2=elastic, 4=inelastic, 16=2n, 51=inelastic to 1st excited level, 102=gamma, 103=proton, 107=alpha. If neutron is a projectile, then 1 stands for (n,tot), 2=(n,n), 4=(n,n'), 16=(n,2n), 51=(n,n1'), 102=(n,gamma), 103=(n,p), 107=(n,alpha).

MF number (Quantity)
MF numbers define quantity, where 1=comment part, 2=neutron resonance parameters, 3=reaction cross section, 4=angular distribution of emitted particles, 5=spectra of emitted particles, 12=photon production, 32=covariances in neutron resonance region, 33=covariances above neutron resonance region.

Product
Final nucleus produced in a reaction. It is rarely used, though it may be userful when looking for activation cross section producing radioactive nucleus in JEFF-3.1/A (Activation) library.

Laboratory
Laboratories where evaluation was done are given in the comment part of each file (target).

Author(s)
Authors of evaluations are given in the comment part of each file. For example, to retrieve all evaluations by P.G. Young one should ask for *Young*.

Examples
Several examples follow in a graphical form. Note: some retrieval and plotting capabilities may be limited, including cross-section plots at higher energies (> 20 MeV); examples were prepared using previous versions of the retrieval system, some changes have been introduced since then.

Example 1: 56Fe(n,g) cross sections, US and Japanese libraries, plot
Example 2: 99Tc(n,n') and (n,2n), cross sections, Japanese library, numerical values and plot
Example 3: All Gd isotopes, (n,tot), cross sections, US library, plot
Example 4: 23Na(n,n) cross sections, US library, plot, compare with experimental data
Example 5: C-natural, (n,n) and (n,n'), angular distribution of emitted neutrons, US library
Example 6: 238U(n,n') and (n,2n), energy spectrum of emitted neutrons, US library, plot
Example 7: 235-U, nu-bar, all libraries
Example 8: 239Pu, decay data, all libraries
Example 9: 235U, individual and cumulative fission product yields, US library, numerical values
Example 10: Selected targets, neutron capture in the pointwise format, US library, zipped numerical data
Example 11: All targets, list of covariance data for neutron resonance parameters, all libraries
Example 12: All targets, thermal neutron scattering, European library
Example 13: 63Cu, full file (material) in original ENDF-6 format, US library

How to?

How to compare evaluated and experimental data
How to download full file (material) in original ENDF-6 format?
How to plot angular distributions?
How to plot energy spectra?
How to plot reactions cross sections?
How to retrieve and compare several reactions?
How to retrieve angular distributions of emitted particles?
How to retrieve average number of neutrons per fission, nu-bar?
How to retrieve covariance data?
How to retrieve cross sections?
How to retrieve decay data?
How to retrieve energy spectra of emitted particles?
How to retrieve fission product yields data?
How to retrieve numerical values of cross sections?
How to retrieve thermal neutron scattering data?
How to retrieve zipped numerical data for selected targets?

What is?

  • What is Library 300° K pointwise?
    It refers to conversion of resonance parameters into cross sections and to Doppler broadening them by taking into account neutron velocities at room temperature. Applies only to the neutron sublibrary in the resonance region. This pointwise version of the library is used for plotting.
  • What is Library basic version?
    It is the library in the original ENDF-6 format, usually with the resonance region given in terms of the resonance parameters. This basic version of the library must be processed by codes such as NJOY to provide suitable input for application codes.
  • What is ENDF-6?
    ENDF-6 is format developed for ENDF/B-VI library. The format is internationally adopted. New ENDF/B-VII library is also stored in this format.
  • What is interpreted ENDF?
    Data stored in the ENDF-6 format are difficult to understand. The interpretted ENDF, developed by R. MacFarlane in Los Alamos, translates ENDF-6 data into human readable format.
  • What is MAT?
    It stands for material (isotope, element). In the ENDF-6 format each material is assigned a unique MAT number. For example, Fe-56 has MAT= 2631.
  • What is nu-bar?
    It is the average number of neutrons per fission.

Why?

  • Why can't I plot low-energy neutron capture cross sections?
    Neutron reactions at low energies (usually below 100 keV) are stored in form of resonance parameters (MF=2). These parameters must be first converted into cross sections by processing that includes Doppler broadening effect usually at 3000 Kelvin. If you retrieved data from the basic library where no processing was applied, cross sections may not be available and thus cannot be plotted.
  • Why can't I retrieve neutron inelastic cross section
    Inelastic neutrons are typically stored under several MT numbers (MT=51,52, ... for transitions to discrete levels, and MT=91 for the continuum part). These partials might be summed up and stored as total inelastic neutron cross section under MT=4 that can be retrieved and plotted as Reaction=(n,inl), Quantity=sig. MT=4, however, is redundant and not mandatory in a basic evaluation. If it is not present, in order to plot inelastic scattering one should sum continuum cross section with cross sections to all discrete levels (this capability is not yet available)

    To retrieve spectrum (Quantity=de) of inelastic neutrons, one should use Reaction=(n,n`) with backward prime to get continuum spectrum.
  • Why can't I retrieve neutron cross-sections for a number of nuclei?
    ENDF libraries contain data for nuclei of technological interest, representing a small fraction (~10%) of all known nuclei.