Nuclear data processing with enhanced version of the legacy NJOY-2016 code.
The NJOY-2016 code is an Open Source code, developed by LANL, and hosted on GitHub code repository. Sandia has made changes to the baseline NJOY-2016 code in order to support applications by the radiation damage community. The application area that our enhancements address are in the development of the energy-dependent radiation response functions for describing radiation damage to materials.
A typical application is iron embrittlement of the critical weld in the pressure vessel for a light water reactor. This energy dependent response permits the research community to correlate observed damage between exposures to the neutron environments in different reactors (typically comparison between commercial reactors and material test reactors). Thus, this enhanced code capability is useful to the academic community and the ASTM reactor pressure vessel embrittlement safety community.
We tried to incorporate our code changes into the current LANL Open Source version of NJOY-2016, but the
NJOY code system has moved on to the NJOY21 code system and there were some compatibility issues with
incorporating our required capability enhancements into the NJOY21 code. LANL has frozen their work on NJOY-2016.
To distinguish our modified version of NJOY from the current LANL GitHub version, we are calling our code
SNL-NJOY-2016 - and have also made our version available to the general public via GitHub.
Our version includes our enhancements to the baseline LANL NJOY-2016 code using the Git software configuration
control system - so configuration control between our version and LANL's version is rigorously maintained.
LANL (Jeremy Conlin) will examine these enhancements and try to find ways to incorporate our changes
into a future LANL version of the NJOY21 code.
"Copyright 2020 National Technology & Engineering Solutions of Sandia, LLC (NTESS). Under the terms of Contract DE-NA0003525 with NTESS, the U.S. Government retains certain rights in this software."
NOTICE:
For five (5) years from 6/11/2020 the United States Government is granted for itself and others acting on its behalf a paid-up, nonexclusive, irrevocable worldwide license in this data to reproduce, prepare derivative works, and perform publicly and display publicly, by or on behalf of the Government. There is provision for the possible extension of the term of this license. Subsequent to that period or any extension granted, the United States Government is granted for itself and others acting on its behalf a paid-up, nonexclusive, irrevocable worldwide license in this data to reproduce, prepare derivative works, distribute copies to the public, perform publicly and display publicly, and to permit others to do so. The specific term of the license can be identified by inquiry made to National Technology and Engineering Solutions of Sandia, LLC or DOE.
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This software has been assigned SCR# 2499.0. This number is an internal tracking number and is a useful reference when contacting the Legal Technology Transfer Center or Licensing and IP Management.
This software was originally assigned Export Control Classification Number (ECCN) EAR 99. However, since this software is to be released as OSS, the software is deemed to be Publicly Available.
Original NJOY README Information Below:
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.
The documentation for NJOY2016 is found in the NJOY2016-manual repository. There, you can find a pre-compiled PDF of the manual.
Instructions for the installation of NJOY2016 are found on our page, Obtaining and Installing NJOY.
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 ofERRORR
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 theCCCC
standard interface files ISOTXS, BRKOXS, and DLAYXS.MATXSR
formats multigroup data for the newerMATXS
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 inACE
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 ofPLOTR
, or special graphics fromHEATR
,COVR
,DTFR
, orACER
, 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 byTHERMR
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 usingGROUPR
, passed toACER
, or displayed usingPLOTR
.
This software is distributed and copyrighted according to the LICENSE file.