Online Catalog
Click on Package Name to get detailed information.
Click on Abstract to read the package abstract.
Click on RSICC Tapelist to view list of files distributed with package.

Note: RESTRICTIONS APPLY TO SOME PACKAGES -
810 -- US DOE 10CFR810 Jurisdiction
FEDC -- US Government Agencies and Their Contractors Only
OECD -- Restricted/See Abstract
RUGA -- Restricted Use Government Authorized
USSO -- US Distribution Only
USUNV -- US Universities Only
Packages with Keyword: MCNP FORMAT
Package NameAbstractRSICC TapelistTitle
ALEPH-LIB-JEFF3.1AbstractD00230 MNYCP 00ACE Format Neutron Cross Section Library based on JEFF3.1.
CRYO-S(A,B)-ACE1AbstractD00253 MNYCP 00Scattering Law and Continuous Energy Cross Section Library of Materials at Cryogenic Temperatures.
FENDL-2.1AbstractD00222 MNYCP 00Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications.
FSX96AbstractD00190 MNYWS 00Collection of Continuous Energy Cross Section Libraries for MCNP Based on JENDL 3.2, JENDL, Fusion File and Dosimetry File.
FSXLIB-J3AbstractD00165 ALLCP 00MCNP continuous energy neutron cross section library based on JENDL-3.
FSXLIB-J33AbstractD00223 MNYCP 01Continuous Energy Neutron Cross Section Library for MCNP Based on JENDL 3.3.
IRDF-2002AbstractD00229 MNYCP 01The International Reactor Dosimetry File.
MCB63NEA.BOLIBAbstractD00216 MNYCP 00ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code.
MCJEF22NEA.BOLIBAbstractD00203 MNYCP 01JEF 2.2 Cross Section Library for the MCNP Monte Carlo Code.
MCJEFF3.1NEAAbstractD00228 MNYCP 00Neutron Cross Section Library Based on JEFF3.1 for Use with MCNP.
TENDL-2008-ACEAbstractD00243 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2010-ACEAbstractD00248 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2011-ACEAbstractD00252 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2012-ACEAbstractD00266 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TSL-ACE/2013AbstractD00270 ALLCP 00TSL-ACE/2013
UTXS6AbstractD00211 MNYCP 00MCNP Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1365K.
VIP-MANAbstractD00256 MNYCP 00Computational Phantom.
YUMMYAbstractD00221 MNYCP 00Multi-temperature, Neutron Cross Section Library Based on ENDF/B-V and ENDF/B-VI for use with MCNP.
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The Radiation Safety Information Computational Center (RSICC) collects, analyzes, maintains, and distributes software in the areas of radiation transport and safety. RSICC resides in the Nuclear Energy and Fuel Cycle Division (NEFCD) at Oak Ridge National Laboratory.