ENDF, NJOY, and Applications

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/B-III, ENDF/B-IV, and ENDF/B-VI, each version adding both improved data and new capabilities for representing nuclear data . The current ENDF-6 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 web-based 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 (energy-dependent) cross sections from ENDF resonance parameters and interpolation schemes.

  • BROADR Doppler broadens and thins pointwise cross sections.

  • UNRESR computes effective self-shielded pointwise cross sections in the unresolved energy range.

  • HEATR generates pointwise heat production cross sections (KERMA coefficients) and radiation-damage cross sections.

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

  • GROUPR generates self-shielded multigroup cross sections, group-to-group scattering matrices, photon-production matrices, and charged-particle cross sections from pointwise input.

  • GAMINR calculates multigroup photoatomic cross sections, KERMA coefficients, and group-to-group 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 blocked-binary format.

  • DTFR formats multigroup data for transport codes that accept formats based in the DTF-IV code.

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

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

  • RESXSR prepares pointwise cross sections in a CCCC-like form for thermal flux calculators.

  • ACER prepares libraries in ACE format for the Los Alamos continuous-energy Monte Carlo code MCNP.

  • POWR prepares libraries for the EPRI-CELL and EPRI-CPM codes.

  • WIMSR prepares libraries for the thermal reactor assembly codes WIMS-D and WIMS-E.

  • PLOTR reads ENDF-format 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 unresolved-resonance probability tables for use in representing resonance self-shielding effects in the MCNP Monte Carlo code.

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

  • GASPR generates gas-production 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 pages--just 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 thermal-reactor 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.


23 January 2013 T-2 Nuclear Information Service ryxm@lanl.gov