Online Catalog
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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
USSO -- US Distribution Only
USUNV -- US Universities Only
Packages starting with F
Package NameAbstractRSICC TapelistTitle
F5TABAbstractP00221 D0780 00Code System for Converting Energy Distribution Cross Section Data to Tabulated Data.
FAMRECAbstractP00167 C7600 01Fuel Assembly Mechanical Response Code System.
FANACAbstractP00179 I3033 00A Shape Analysis Code Package for Resonance Parameter Extraction from Neutron Capture Data for Light- and Medium-Weight Nuclei.
FANALAbstractP00178 I3033 00A Least-Squares Shape Analysis Code System.
FANGAbstractP00140 C0000 00An Angular Folding Code System for Channel Theory Analysis.
FANGAbstractP00140 I0360 00An Angular Folding Code System for Channel Theory Analysis.
FANTOMAbstractC00375 BESM6 00Monte Carlo Calculation of the Response of an External Detector to a Photon Source in the Lungs of a Heterogeneous Phantom.
FASTER IIIAbstractC00168 U1108 00Monte Carlo Neutron and Photon Transport Code System in Complex Geometries.
FASTER-IIIAbstractC00168 I3675 00Monte Carlo Neutron and Photon Transport Code System in Complex Geometries.
FASTGRASSAbstractP00479 MNYCP 00Code System to Predict Fission Product Release in Ubase Fuels.
FASTPLOT 1.0AbstractP00354 IBMPC 00Interface to Microsoft FORTRAN Graphics.
FATDUDAbstractP00080 I0360 00Foil Activation Data Unfolding Code System.
FBSAMAbstractP00103 I0360 00User-Storage - Magnetic Disk Data Manipulator.
FCXSECAbstractD00085 PC386 0122 Neutron, 21 Gamma-Ray Group Cross Section Libraries in ANISN Format for Nuclear Fuel Cycle Shielding Calculations.
FDKRAbstractC00541 I4381 00Radioactivity and Dose Rate Calculation Code for Fission, Fusion and Hybrid Reactors.
FDMXPCAbstractP00322 IPCAT 00Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format.
FE3DGWAbstractC00531 D0780 00Code System for Finite-Element, Three-Dimensional Ground-Water Flow Analysis.
FEAST METALAbstractP00563 MNYCP 00Fuel Engineering and Structural Analysis Tool.
FEDGROUP-3AbstractP00123 I0360 00Program System for Processing Evaluated Nuclear Data in ENDF/B, KEDAK or UKNDL Format to Constants to be Used in Reactor Physics Calculation.
FEDGROUPC86REV3AbstractP00194 MNYCP 01Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation.
FEDGROUP-RAbstractP00349 MNYCP 00Multigroup Neutron Cross Section Processing System from Data in ENDF/B Format.
FEM-2DAbstractC00260 C6600 00Two-Dimensional Diffusion Theory Code System Based on the Method of Finite Elements.
FEMAXI 6 VER.1AbstractP00536 IBMPC 00Code System for Light Water Reactor Fuel Analysis.
FEMBAbstractC00340 B6700 00A Two-Dimensional Diffusion Theory Finite Element Program.
FEMRZAbstractC00342 F2307 00A Finite-Element Method Two-Dimensional Multigroup Neutron Transport Code System, (r,z) Geometry.
FEMWASTE/FEMWATERAbstractC00451 C7600 00A Finite-Element Model of Waste and Water Transport through Porous Saturated-Unsaturated Media.
FEMWASTE/FEMWATERAbstractC00451 PC386 00A Finite-Element Model of Waste and Water Transport through Porous Saturated-Unsaturated Media.
FENDL-2.0AbstractD00183 MNYCP 01Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications.
FENDL-2.1AbstractD00222 MNYCP 00Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications.
FEP 4.16AbstractP00440 IBMPC 00Fault-tree, Event tree, & P&ID Editors.
FERDO/FERDAbstractP00102 I3033 00Multichannel Neutron and Gamma-Ray Spectrum Matrix Unfolding Code Systems.
FERDORAbstractP00017 I7090 00Spectra Unfolding Codes.
FERDORAbstractP00017 U1108 00Spectra Unfolding Codes.
FERD-PCAbstractP00273 IBMPC 00Interactive Multichannel Neutron and Gamma-Ray Spectrum Matrix Unfolding Code System.
FERRETAbstractP00145 U0000 00Least-Squares Solution to Nuclear Data and Reactor Physics Problems.
FESHAbstractC00676 CDCMF 00X-Y Multigroup Neutron Transport Code System.
FEWA-FEMAAbstractC00477 I3033 00A Finite Element Model of Water and Other Material through Aquifers.
FEWG1-81AbstractD00031 I0370 06Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
FEWG1-85AbstractD00031 I0360 07Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
FGR-DOSEAbstractD00167 ALLCP 01Dose Coefficients from Federal Guidance Reports 11 and 12.
FGXRRSAbstractD00132 C0000 00Few Group Cross Section Library for Research Reactor Calculations.
FIGEROAbstractP00149 C0000 00Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations.
FINELMAbstractC00483 MFMWS 00Multigroup Finite Element Diffusion Code System.
FIPDIGAbstractC00251 I0360 00One-Dimensional Time-Dependent Fission Product Diffusion Code System.
FIRACAbstractP00444 CY000 00Nuclear Facilities Fire Accident Model
FIREDATAAbstractD00125 PC486 00Nuclear Power Plant Fire Data Base for Personal Computers.
FISP-6AbstractC00538 I3090 00An Enhanced Code for the Evaluation of Fission Product Inventories and Decay Heat.
FISPINAbstractC00413 ICL00 00Nuclide Inventory Calculation System.
FIS-PRODAbstractD00152 ALLCP 00Chinese Evaluated Fission Product Yield Library in ENDF/B-V Format.
FISSP & CLOUDAbstractC00163 MNYCP 01Fission Product Inventory, Release, Transport and Dose Calculation.
FITOCOAbstractP00189 C0175 00Converter of Fine-Group Flux Density and Cross Section Data to Coarse Group Values.
FLANGE-ORNLAbstractP00566 I0360 00Flanged Pipe Joint Stress Analysis, Internal Pressure, Moment Loads, Temperature.
FLEPAbstractD00022 I3033 00Coefficients for the Analytic Representation of Nonelastic Cross Sections and Particle-Emission Spectra from Various Nucleon-Nucleus Collisions in the Energy Range 25 to 400 MeV.
FLODISAbstractP00417 I0360 00Code System to Calculate Thermal Response of FSV HTGR Core.
FLOWPLOT IIAbstractP00234 I3033 00Fluid Dynamics and Heat Transfer Plotting Package.
FLUKA05-PRE-LIBAbstractD00260 PCX86 00FLUKA05 Multi-Group, Multi-Purpose Nuclear Data Library, Neutrons, Photons, Charged Particles.
FLUKA-TRANKAAbstractC00207 C6600 00Three-Dimensional High-Energy Extranuclear Hadron Cascade Monte Carlo System for Cylindrical Backstop Geometries.
FLUNGAbstractD00086 I3033 00Coupled 35-Group Neutron and 21-Group Gamma Ray, P3 Cross Sections for Fusion Applications.
FLUSHAbstractP00043 C6600 00Spectral Unfolding Code - Stepwise Regression of System Response Functions.
FLYSPEC-SHORTSAbstractP00196 C7600 00Neutron Unfolding Code System for Reducing Proton-Recoil Pulse-Height Obtained with NE-213 Liquid Scintillator.
FOCUSAbstractC00390 I3033 00Adjoint Monte Carlo Neutron Transport Code System.
FONTAAbstractC00423 S4044 00Code System For Calculating Individual And Collective Doses From Reactor Accidents Using Pasquill's Plume Model.
FOODAbstractC00403 U1108 00Calculation of Radiation Dose to Man from Radionuclides in the Environment.
FORECAST V3.0AbstractP00384 IBMPC 00Forecast Regulatory Effects Cost Analysis Program.
FORISTAbstractP00092 C0000 00Neutron Spectrum Unfolding Code System - Iterative Smoothing Technique.
FORISTAbstractP00092 I0360 00Neutron Spectrum Unfolding Code System - Iterative Smoothing Technique.
FORSENAbstractP00170 I0360 00A Multigroup Processing Code for Use with Sensitivity Profiles to Assess the Effect of Cross Section Changes.
FORSIM VIAbstractP00078 C6600 00A Fortran-Oriented Simulation Package for the Automated Solution of Partial and Ordinary Differential Equation Systems.
FORSSAbstractC00334 C0000 00A Sensitivity and Uncertainty Analysis Code System.
FORSSAbstractC00334 I0360 00A Sensitivity and Uncertainty Analysis Code System.
FOTELP-2K6AbstractC00581 MNYCP 03Monte Carlo Simulation of Photons, Electrons and Positrons Transport.
FOURACESAbstractP00183 I0370 00Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries.
FPDLAbstractD00066 I0360 00Fission Product Yields, Gamma Ray and Beta Spectra in ENDF-III Format for 235U, 238U, 239Pu, 232Th, and 233U.
FPGAMAbstractC00386 F2307 00Calculation of Fission-Product Gamma-Ray Spectra.
FPICAbstractC00028 I3675 00Fission Product Inventory Code.
FPIPAbstractC00162 C6600 00Fission Product Inventory Code System.
FPZDAbstractC00603 PC386 00Code System for Multigroup Neutron Diffusion/Depletion Calculations.
FRANCOAbstractP00363 MNYCP 00Finite Element Fuel Rod Analysis Code System.
FRANTIC3AbstractP00406 CDCMF 00Time-Dependent Reliability Analysis.
FRAPCON2
USSO
AbstractP00517 MFMWS 00Fuel Rod Thermal-Mechanical Behavior, Versions FRAPCON2, FRAPCON2/VIM4, & FRAPCON2/VIM5.
FRAPT6/MOD1
USSO
AbstractP00436 C0176 00Code System for Transient Analysis of Fuel Rods.
FRAPT6/V21
USSO
AbstractP00436 C0176 01Code System for Transient Analysis of Fuel Rods.
FRCRL2AbstractC00231 C6400 00Calculation of Fission-Product Release in Reactor Accident Analyses.
FREEFORMAbstractP00081 I0360 00Free-Form Input Reading Routines.
FSCATTAbstractC00186 I3033 00Discrete Ordinates Gamma-Ray Transport Code System in Plane Geometry.
FSCATTAbstractC00186 U1108 00Discrete Ordinates Gamma-Ray Transport Code System in Plane Geometry.
FSKY4CAbstractC00771 PCX86 00Gamma Ray Skyshine Analysis Code.
FSX96AbstractD00190 MNYWS 00Collection of Continuous Energy Cross Section Libraries for MCNP Based on JENDL 3.2, JENDL, Fusion File and Dosimetry File.
FSXJ32AbstractD00244 MNYCP 00A Continuous Energy Cross Section MCNP Nuclear Data Library Based on JENDL-3.2.
FSXLIB-J3AbstractD00165 ALLCP 00MCNP continuous energy neutron cross section library based on JENDL-3. See DLC-190/FSX96 based on JENDL3.2.
FSXLIB-J33AbstractD00223 MNYCP 01Continuous Energy Neutron Cross Section Library for MCNP Based on JENDL 3.3.
FTFAbstractD00056 I0360 00Multigroup Neutron and Gamma-Ray Dose Transmission Factors for Concrete Slabs.
FUELSDATAAbstractP00446 C7600 00Code System to Model Verification Fuel Rod Data.
FURNACEAbstractC00615 C0740 00Code System for Neutronic Calculations in Three Dimension Toroidal Geometry.
The Radiation Safety Information Computational Center (RSICC) is a Department of Energy Specialized Information Analysis Center (SIAC) authorized to collect, analyze, maintain, and distribute computer software and data sets in the areas of radiation transport and safety. RSICC resides in the Reactor and Nuclear Systems Division (RNSD) at Oak Ridge National Laboratory.