| Packages starting with F |
| Package Name | Abstract | RSICC Tapelist | Title |
| F5TAB | Abstract | P00221 D0780 00 | Code System for Converting Energy Distribution Cross Section Data to Tabulated Data. |
| FAMREC | Abstract | P00167 C7600 01 | Fuel Assembly Mechanical Response Code System. |
| FANAC | Abstract | P00179 I3033 00 | A Shape Analysis Code Package for Resonance Parameter Extraction from Neutron Capture Data for Light- and Medium-Weight Nuclei. |
| FANAL | Abstract | P00178 I3033 00 | A Least-Squares Shape Analysis Code System. |
| FANG | Abstract | P00140 C0000 00 | An Angular Folding Code System for Channel Theory Analysis. |
| FANG | Abstract | P00140 I0360 00 | An Angular Folding Code System for Channel Theory Analysis. |
| FANTOM | Abstract | C00375 BESM6 00 | Monte Carlo Calculation of the Response of an External Detector to a Photon Source in the Lungs of a Heterogeneous Phantom. |
| FASTER III | Abstract | C00168 U1108 00 | Monte Carlo Neutron and Photon Transport Code System in Complex Geometries. |
| FASTER-III | Abstract | C00168 I3675 00 | Monte Carlo Neutron and Photon Transport Code System in Complex Geometries. |
| FASTGRASS | Abstract | P00479 MNYCP 00 | Code System to Predict Fission Product Release in Ubase Fuels. |
| FASTPLOT 1.0 | Abstract | P00354 IBMPC 00 | Interface to Microsoft FORTRAN Graphics. |
| FATDUD | Abstract | P00080 I0360 00 | Foil Activation Data Unfolding Code System. |
| FBSAM | Abstract | P00103 I0360 00 | User-Storage - Magnetic Disk Data Manipulator. |
| FCXSEC | Abstract | D00085 PC386 01 | 22 Neutron, 21 Gamma-Ray Group Cross Section Libraries in ANISN Format for Nuclear Fuel Cycle Shielding Calculations. |
| FDKR | Abstract | C00541 I4381 00 | Radioactivity and Dose Rate Calculation Code for Fission, Fusion and Hybrid Reactors. |
| FDMXPC | Abstract | P00322 IPCAT 00 | Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format. |
| FE3DGW | Abstract | C00531 D0780 00 | Code System for Finite-Element, Three-Dimensional Ground-Water Flow Analysis. |
| FEAST METAL | Abstract | P00563 MNYCP 00 | Fuel Engineering and Structural Analysis Tool. |
| FEDGROUP-3 | Abstract | P00123 I0360 00 | Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation. |
| FEDGROUPC86REV3 | Abstract | P00194 MNYCP 01 | Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation. |
| FEDGROUP-R | Abstract | P00349 MNYCP 00 | Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation. |
| FEM-2D | Abstract | C00260 C6600 00 | Two-Dimensional Diffusion Theory Code System Based on the Method of Finite Elements. |
| FEMAXI 6 VER.1 | Abstract | P00536 IBMPC 00 | Code System for Light Water Reactor Fuel Analysis. |
| FEMB | Abstract | C00340 B6700 00 | A Two-Dimensional Diffusion Theory Finite Element Program. |
| FEMRZ | Abstract | C00342 F2307 00 | A Finite-Element Method Two-Dimensional Multigroup Neutron Transport Code System, (r,z) Geometry. |
| FEMWASTE/FEMWATER | Abstract | C00451 C7600 00 | A Finite-Element Model of Waste and Water Transport through Porous Saturated-Unsaturated Media. |
| FEMWASTE/FEMWATER | Abstract | C00451 PC386 00 | A Finite-Element Model of Waste and Water Transport through Porous Saturated-Unsaturated Media. |
| FENDL-2.0 | Abstract | D00183 MNYCP 01 | Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications. |
| FENDL-2.1 | Abstract | D00222 MNYCP 00 | Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications. |
| FEP 4.16 | Abstract | P00440 IBMPC 00 | Fault-tree, Event tree, & P&ID Editors. |
| FERDO/FERD | Abstract | P00102 I3033 00 | Multichannel Neutron and Gamma-Ray Spectrum Matrix Unfolding Code Systems. |
| FERDOR | Abstract | P00017 I7090 00 | Spectra Unfolding Codes. |
| FERDOR | Abstract | P00017 U1108 00 | Spectra Unfolding Codes. |
| FERD-PC | Abstract | P00273 IBMPC 00 | Interactive Multichannel Neutron and Gamma-Ray Spectrum Matrix Unfolding Code System. |
| FERRET | Abstract | P00145 U0000 00 | Least-Squares Solution to Nuclear Data and Reactor Physics Problems. |
| FESH | Abstract | C00676 CDCMF 00 | X-Y Multigroup Neutron Transport Code System. |
| FEWA-FEMA | Abstract | C00477 I3033 00 | A Finite Element Model of Water and Other Material through Aquifers. |
| FEWG1-81 | Abstract | D00031 I0370 06 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| FEWG1-85 | Abstract | D00031 I0360 07 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| FGR-DOSE | Abstract | D00167 ALLCP 01 | Dose Coefficients from Federal Guidance Reports 11 and 12. |
| FGXRRS | Abstract | D00132 C0000 00 | Few Group Cross Section Library for Research Reactor Calculations. |
| FIGERO | Abstract | P00149 C0000 00 | Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations. |
| FINELM | Abstract | C00483 MFMWS 00 | Multigroup Finite Element Diffusion Code System. |
| FIPDIG | Abstract | C00251 I0360 00 | One-Dimensional Time-Dependent Fission Product Diffusion Code System. |
| FIRAC | Abstract | P00444 CY000 00 | Nuclear Facilities Fire Accident Model |
| FIREDATA | Abstract | D00125 PC486 00 | Nuclear Power Plant Fire Data Base for Personal Computers. |
| FISP-6 | Abstract | C00538 I3090 00 | An Enhanced Code for the Evaluation of Fission Product Inventories and Decay Heat. |
| FISPACT-II 5.0 | Abstract | C00836 MNYCP 03 | Inventory Simulation Platform for Nuclear Observables and Materials Science. |
| FISPIN | Abstract | C00413 ICL00 00 | Nuclide Inventory Calculation System. |
| FIS-PROD | Abstract | D00152 ALLCP 00 | Chinese Evaluated Fission Product Yield Library in ENDF/B-V Format. |
| FISSP & CLOUD | Abstract | C00163 MNYCP 01 | Fission Product Inventory, Release, Transport and Dose Calculation. |
| FITOCO | Abstract | P00189 C0175 00 | Converter of Fine-Group Flux Density and Cross Section Data to Coarse Group Values. |
| FLANGE-ORNL | Abstract | P00566 I0360 00 | Flanged Pipe Joint Stress Analysis, Internal Pressure, Moment Loads, Temperature. |
| FLEP | Abstract | D00022 I3033 00 | Coefficients 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. |
| FLODIS | Abstract | P00417 I0360 00 | Code System to Calculate Thermal Response of FSV HTGR Core. |
| FLOWPLOT II | Abstract | P00234 I3033 00 | Fluid Dynamics and Heat Transfer Plotting Package. |
| FLUKA05-PRE-LIB | Abstract | D00260 PCX86 00 | FLUKA05 Multi-Group, Multi-Purpose Nuclear Data Library, Neutrons, Photons, Charged Particles. |
| FLUKA-TRANKA | Abstract | C00207 C6600 00 | Three-Dimensional High-Energy Extranuclear Hadron Cascade Monte Carlo System for Cylindrical Backstop Geometries. |
| FLUNG | Abstract | D00086 I3033 00 | Coupled 35-Group Neutron and 21-Group Gamma Ray, P3 Cross Sections for Fusion Applications. |
| FLUSH | Abstract | P00043 C6600 00 | Spectral Unfolding Code - Stepwise Regression of System Response Functions. |
| FLYSPEC-SHORTS | Abstract | P00196 C7600 00 | Neutron Unfolding Code System for Reducing Proton-Recoil Pulse-Height Obtained with NE-213 Liquid Scintillator. |
| FOCUS | Abstract | C00390 I3033 00 | Adjoint Monte Carlo Neutron Transport Code System. |
| FONTA | Abstract | C00423 S4044 00 | Code System For Calculating Individual And Collective Doses From Reactor Accidents Using Pasquill's Plume Model. |
| FOOD | Abstract | C00403 U1108 00 | Calculation of Radiation Dose to Man from Radionuclides in the Environment. |
| FORECAST V3.0 | Abstract | P00384 IBMPC 00 | Forecast Regulatory Effects Cost Analysis Program. |
| FORIST | Abstract | P00092 C0000 00 | Neutron Spectrum Unfolding Code System - Iterative Smoothing Technique. |
| FORIST | Abstract | P00092 I0360 00 | Neutron Spectrum Unfolding Code System - Iterative Smoothing Technique. |
| FORSEN | Abstract | P00170 I0360 00 | A Multigroup Processing Code for Use with Sensitivity Profiles to Assess the Effect of Cross Section Changes. |
| FORSIM VI | Abstract | P00078 C6600 00 | A Fortran-Oriented Simulation Package for the Automated Solution of Partial and Ordinary Differential Equation Systems. |
| FORSS | Abstract | C00334 C0000 00 | A Sensitivity and Uncertainty Analysis Code System. |
| FORSS | Abstract | C00334 I0360 00 | A Sensitivity and Uncertainty Analysis Code System. |
| FOTELP-2014 | Abstract | C00581 MNYCP 04 | Monte Carlo Simulation of Photons, Electrons and Positrons Transport. |
| FOURACES | Abstract | P00183 I0370 00 | Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries. |
| FPDL | Abstract | D00066 I0360 00 | Fission Product Yields, Gamma Ray and Beta Spectra in ENDF-III Format for 235U, 238U, 239Pu, 232Th, and 233U. |
| FPGAM | Abstract | C00386 F2307 00 | Calculation of Fission-Product Gamma-Ray Spectra. |
| FPIC | Abstract | C00028 I3675 00 | Fission Product Inventory Code. |
| FPIP | Abstract | C00162 C6600 00 | Fission Product Inventory Code System. |
| FPZD | Abstract | C00603 PC386 00 | Code System for Multigroup Neutron Diffusion/Depletion Calculations. |
| FRANCO | Abstract | P00363 MNYCP 00 | Finite Element Fuel Rod Analysis Code System. |
| FRANTIC3 | Abstract | P00406 CDCMF 00 | Time-Dependent Reliability Analysis. |
| FRAPCON2 | Abstract | P00517 MFMWS 00 | Fuel Rod Thermal-Mechanical Behavior. |
FRAPT6/MOD1
USSO | Abstract | P00436 C0176 00 | Code System for Transient Analysis of Fuel Rods. |
FRAPT6/V21
USSO | Abstract | P00436 C0176 01 | Code System for Transient Analysis of Fuel Rods. |
| FRCRL2 | Abstract | C00231 C6400 00 | Calculation of Fission-Product Release in Reactor Accident Analyses. |
| FREEFORM | Abstract | P00081 I0360 00 | Free-Form Input Reading Routines. |
| FSCATT | Abstract | C00186 I3033 00 | Discrete Ordinates Gamma-Ray Transport Code System in Plane Geometry. |
| FSCATT | Abstract | C00186 U1108 00 | Discrete Ordinates Gamma-Ray Transport Code System in Plane Geometry. |
| F-SCORE | Abstract | P00617 PCX86 00 | F-Score Nuclide ID Scoring Applications |
| FSKY4C | Abstract | C00771 PCX86 00 | Gamma Ray Skyshine Analysis Code. |
| FSX96 | Abstract | D00190 MNYWS 00 | Collection of Continuous Energy Cross Section Libraries for MCNP Based on JENDL 3.2, JENDL, Fusion File and Dosimetry File. |
| FSXJ32 | Abstract | D00244 MNYCP 00 | A Continuous Energy Cross Section MCNP Nuclear Data Library Based on JENDL-3.2. |
| FSXLIB-J3 | Abstract | D00165 ALLCP 00 | MCNP continuous energy neutron cross section library based on JENDL-3. |
| FSXLIB-J33 | Abstract | D00223 MNYCP 01 | Continuous Energy Neutron Cross Section Library for MCNP Based on JENDL 3.3. |
| FTF | Abstract | D00056 I0360 00 | Multigroup Neutron and Gamma-Ray Dose Transmission Factors for Concrete Slabs. |
| FUELSDATA | Abstract | P00446 C7600 00 | Code System to Model Verification Fuel Rod Data. |
| FURNACE | Abstract | C00615 C0740 00 | Code System for Neutronic Calculations in Three Dimension Toroidal Geometry. |