
The ENDF formats were originally developed for use in the
US national nuclear data files called ENDF/B (the Evaluated
Nuclear Data Files). These files went through various versions
with names like ENDF/BIII, ENDF/BIV, and ENDF/BVI, each
version adding both improved data and new capabilities for
representing nuclear data . The current ENDF6 format can
represent cross sections for neutrons, photons, and charged
particles, including particle yields and distributions in angle
and energy, for energies up to several hundred MeV, the
radioactive decay properties of reaction products, and
estimated errors and covariances of the various nuclear
parameters. The ENDF format is now widely used around the
world, including the JEF files in Europe, the JENDL files in
Japan, and the BROND files in Russia. Thus, even though NJOY
was originally designed to work with the US ENDF/B libraries,
it now provides a universal capability to work with nuclear
data libraries all over the world. For more information on
the ENDF format, look at the
Introduction to the ENDF Format, which is another webbased
course similar to this one.
The NJOY Nuclear Data Processing System is a modular computer
code designed to read evaluated data in ENDF format, transform
the data in various ways, and output the results as libraries
designed to be used in various applications. Each module performs
a well defined processing task. The modules are essentially
independent programs, and they communicate with each other using
input and output files, plus a very few common variables.

NJOY directs the flow of data through the other modules
and contains a library of common functions and subroutines
used by the other modules.

RECONR reconstructs pointwise (energydependent) cross
sections from ENDF resonance parameters and interpolation
schemes.

BROADR Doppler broadens and thins pointwise cross
sections.

UNRESR computes effective selfshielded pointwise cross
sections in the unresolved energy range.

HEATR generates pointwise heat production cross sections
(KERMA coefficients) and radiationdamage cross sections.

THERMR produces cross sections and energytoenergy
matrices for free or bound scatterers in the thermal energy
range.

GROUPR generates selfshielded multigroup cross sections,
grouptogroup scattering matrices, photonproduction matrices,
and chargedparticle cross sections from pointwise input.

GAMINR calculates multigroup photoatomic cross sections,
KERMA coefficients, and grouptogroup photon scattering
matrices.

ERRORR computes multigroup covariance matrices from ENDF
uncertainties.

COVR reads the output of ERRORR and performs covariance
plotting and output formatting operations.

MODER converts ENDF "tapes" back and forth between
ASCII format and the special NJOY blockedbinary format.

DTFR formats multigroup data for transport codes that
accept formats based in the DTFIV code.

CCCCR formats multigroup data for the CCCC standard
interface files ISOTXS, BRKOXS, and DLAYXS.

MATXSR formats multigroup data for the newer MATXS
material crosssection interface file, which works with the
TRANSX code to make libraries for many particle transport
codes.

RESXSR prepares pointwise cross sections in a CCCClike
form for thermal flux calculators.

ACER prepares libraries in ACE format for the Los Alamos
continuousenergy Monte Carlo code MCNP.

POWR prepares libraries for the EPRICELL and EPRICPM
codes.

WIMSR prepares libraries for the thermal reactor assembly
codes WIMSD and WIMSE.

PLOTR reads ENDFformat files and prepares plots of
cross sections or perspective views of distributions for
output using VIEWR.

VIEWR takes the output of PLOTR, or special graphics from
HEATR, COVR, DTFR, or ACER, and converts the plots into
Postscript format for printing or screen display.

MIXR is used to combine cross sections into elements or
other mixtures, mainly for plotting.

PURR generates unresolvedresonance probability tables for
use in representing resonance selfshielding effects in the MCNP
Monte Carlo code.

LEAPR generates ENDF scatteringlaw files (File 7) for
moderator materials in the thermal range. These scatteringlaw
files can be used by THERMR to produce the corresponding cross
sections.

GASPR generates gasproduction cross sections in pointwise
format from basic reaction data in an ENDF evaluation. These
results can be converted to multigroup form using GROUPR,
passed to ACER, or displayed using PLOTR.
More details on all of these modules will be found on subsequent
pagesjust keep pressing NEXT, or
to browse through the pages in other orders, use the
INDEX links.
Note that some of these modules do calculations to transform the
evaluated nuclear data, and others format the results of the
calculations for various nuclear applications. One very basic
application is the multigroup particle transport code, which is
used to compute neutron and photon fluxes and reaction rates for
reactor design, shielding and radiation protection, criticality
safety, experimental facility design, medical applications, and
so on. Familiar examples of such codes include ANISN from Oak
Ridge and the "DANT" series from Los Alamos. NJOY can prepare
data for these codes through several paths, including DTFR,
CCCCR, and MATXSR. These days, people are making more and more
use of the powerful Monte Carlo method, which uses very detailed
and faithful representations of complex problems. Data for
the popular MCNP code can be produced with ACER. For direct
thermalreactor core calculations, the WIMS code series from
Great Britain is very widely used, and NJOY's WIMSR module can
produce the appropriate libraries for it. The modular structure
of NJOY lends itself to adding additional applications without
having to reinvent the core transformations that NJOY does to
evaluated nuclear data.
