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: ENDF/B FORMAT
Package NameAbstractRSICC TapelistTitle
ACOHAbstractC00191 I3675 00Aerojet COHORT Monte Carlo Code System.
BCGAbstractC00578 C0170 00A Code For Calculating Pointwise Neutron Spectra and Criticality in Fast Reactor Cells.
BOXERAbstractC00766 MNYWS 00Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations
ENDVER/GUIAbstractP00572 PCX86 00The ENDF File Verification Support Package.
EPRI-CINDERAbstractC00309 C6600 00General Point-Depletion and Fission Product Code System and Four-Group Fission Product Neutron Absorption Chain Data Library Generated from ENDF/B-IV for Thermal Reactors.
ESPAbstractC00193 I0360 00General Purpose Monte Carlo Neutron Transport Code System.
FISPINAbstractC00413 ICL00 00Nuclide Inventory Calculation System.
FOCUSAbstractC00390 I3033 00Adjoint Monte Carlo Neutron Transport Code System.
O6RAbstractC00128 I3675 00A General-Purpose Monte Carlo Transport Code System.
PADF-2007AbstractD00259 PCX86 00Proton Activation Data File in ENDF-6 Format.
PALLAS-1D(VII)AbstractC00380 FM380 00Multigroup Time-Independent Neutron Transport Code System for Plane or Spherical Geometry.
PLOTENDFAbstractP00214 I3033 00A Program for Producing Graphical Output.
PREPRO2019AbstractP00351 MNYCP 10Pre-Processing Code System for Data in ENDF/B Format.
RMET21AbstractC00597 D0VAX 00Detailed Space and Energy Treatment of Neutron Resonances for Homogeneous Mixtures and Cylinderized Reactor Cells.
SAM-CEAbstractC00187 C6600 00Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations.
SAM-CEAbstractC00187 I0360 00Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations.
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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.