| Packages with Subject: NUCLEAR DATA GENERATION FOR TRANSPORT CODE INPUT |
| Package Name | Abstract | RSICC Tapelist | Title |
| 1DX | Abstract | P00096 U1108 00 | A One-Dimensional Diffusion Code System for Producing Energy Group Collapsed and Self-Shielded Cross Sections. |
| ABLEIT-TRANS | Abstract | P00247 C0175 00 | Error Propagation Analysis for Burnup Calculation. |
| ADEFTA 4.1 | Abstract | P00543 MNYCP 01 | Atomic Densities for Transport Analysis Script. |
| ADLER III | Abstract | P00058 I0360 00 | A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters. |
| ALEPH-LIB-JEFF3.1 | Abstract | D00230 MNYCP 00 | ACE Format Neutron Cross Section Library based on JEFF3.1. |
| AMARA | Abstract | P00079 I3675 00 | Nuclear Data Adjustment Using Lagrange's Multipliers Method. |
| AREAD | Abstract | P00088 I0360 00 | Input Data Processor for Transport Codes. |
| AUTOJOM-JOMREAD | Abstract | P00008 C6600 00 | Computer Programs to Generate or Check Coefficients for Quadratic Equations Describing 3D Geometries. |
| BOT3P-5.3 | Abstract | P00530 MNYCP 02 | Code System for 2D and 3D Mesh Generation and Graphical Display of Geometry and Results for Radiation Transport Codes. |
| BUCORST | Abstract | P00339 PC386 00 | A Code to Prepare Burnup-Dependent Multigroup Nuclear Reactor Source Terms. |
CALENDF-2010
OECD | Abstract | P00578 PCX86 00 | Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations. |
| CASTHY | Abstract | P00316 FM000 00 | Statistical Model Calculation for Neutron Cross Sections and Capture Gamma-Ray Spectra. |
| CEPXS/ONELD 1.0 | Abstract | C00544 MNYCP 02 | One-Dimensional Coupled Electron-Photon Multigroup Discrete Ordinates Code System. |
| CHENDF 7.02 | Abstract | P00333 MNYCP 05 | Codes for Handling ENDF/B-V and ENDF/B-VI Data. |
| CNCSN 2009 | Abstract | C00726 PCX86 01 | One, Two- and Three-Dimensional Coupled Neutral and Charged Particle SN Parallel Multi-Threaded Code System. |
| COAG-II | Abstract | P00070 I0360 00 | Calculation of the Westcott Epithermal Index and the Westcott 2200 m/s Neutron Flux. |
| CODAC (2) | Abstract | P00073 I0360 00 | For TIMOC 72, Monte Carlo Three-Dimensional Neutron Transport Code's Data Generator. |
| COMAND | Abstract | P00091 I0360 00 | A Multigroup ANISN Cross Section Data Library Collapsing Code System. |
| COMBINE-PC | Abstract | P00286 IBMPC 00 | Code System to Compute Neutron Spectra and ENDF/B Version 5 Based Multigroup Neutron Constants. |
| COMPAR | Abstract | P00240 C0170 00 | Compares Multigroup Cross Sections Generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS. |
| COMPLOT | Abstract | P00259 IBMMF 00 | Convert EXFOR Format Data to Computation Format and Plot Comparisons of EXFOR and ENDF/B Evaluated Data (Version 86-1). |
| CRECTJ5 | Abstract | P00250 D0780 00 | A Computer Program for Compilation of Evaluated Nuclear Data in ENDF/B Format. |
| CRESO | Abstract | P00184 I3081 00 | Resonance Data-Handling Code System. |
| DANCOFF3 | Abstract | P00279 D8810 00 | Calculates Dancoff Correction. |
| DASQHE | Abstract | P00278 D8810 00 | Calculates Dancoff Corrections Factors. |
| DATINIT | Abstract | P00258 DGMV1 00 | Interactive Program To Access Photon Interaction Data. |
| DEPLETOR | Abstract | P00523 MNYCP 00 | Code System to Provide Depletion Capability to the U.S. NRC PARCS Code |
| DETAN 95 | Abstract | P00361 MNYCP 00 | Code System to Calculate Spectrum-Averaged Cross Sections and Detector Responses in Neutron Spectra. |
| DINT | Abstract | P00049 C6600 00 | Multigroup Coherent-Incoherent Cross Section Data Generator for Photon Transport Calculations. |
| DINT | Abstract | P00049 I0360 00 | Multigroup Coherent-Incoherent Cross Section Data Generator for Photon Transport Calculations. |
| DOQDP | Abstract | P00110 I0360 00 | Discrete Ordinates Quadrature Generator. |
| DSNQUAD | Abstract | P00251 IPCXT 00 | Calculates Angular Quadrature Weights and Cosines. |
| EDISTR | Abstract | P00191 I3033 00 | Prepares a Nuclear Decay Data Base for Internal Radiation Dosimetry Calculations. |
| EMPIRE-II | Abstract | P00497 PC586 01 | Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections. |
| ENBAL2 | Abstract | P00160 I0370 00 | A Program to Generate Multigroup Neutron Kerma Factors. |
| EPRI-CINDER | Abstract | C00309 C6600 00 | General 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. |
| ERIC-2 | Abstract | P00119 I0360 00 | Calculator of Resonance Integral and Effective Capture and Fission Cross Sections for Fissile and Non-Fissile Nuclides - Thermal or Fast Reactors. |
| ERRORJ | Abstract | P00526 MNYCP 03 | Multigroup Covariance Matrices Generation from ENDF/B-6 Format. |
| ESTIMA | Abstract | P00201 I3033 00 | A Code System for Calculating Average Parameters from Sets of Resolved Resonance Parameters. |
| ETHEL | Abstract | P00217 I0360 00 | Code System for Generating Cross Sections for PSR-128/THERMOS. |
| EURCYL | Abstract | P00076 I0370 00 | Finite Element Three-Dimensional Mesh Generator for Cylinder - Cylinder Intersections. |
| EVAP | Abstract | P00010 I0360 00 | Calculation of Particle Evaporation from Excited Compound Nuclei. |
| FDMXPC | Abstract | P00322 IPCAT 00 | Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format. |
| FEDGROUPC86REV3 | Abstract | P00194 MNYCP 01 | Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation. |
| FIGERO | Abstract | P00149 C0000 00 | Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations. |
| FITOCO | Abstract | P00189 C0175 00 | Converter of Fine-Group Flux Density and Cross Section Data to Coarse Group Values. |
| FLYSPEC-SHORTS | Abstract | P00196 C7600 00 | Neutron Unfolding Code System for Reducing Proton-Recoil Pulse-Height Obtained with NE-213 Liquid Scintillator. |
| FORSEN | Abstract | P00170 I0360 00 | A Multigroup Processing Code for Use with Sensitivity Profiles to Assess the Effect of Cross Section Changes. |
| FOURACES | Abstract | P00183 I0370 00 | Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries. |
| GALAXY-6 | Abstract | P00098 I0370 00 | Neutron Multigroup Cross Section Processor. |
| GAMLEG-75 | Abstract | P00086 C7600 00 | Multigroup Cross Section Generator for Photon Transport Calculations. |
| GAROL | Abstract | P00033 I7090 00 | Calculation of Resonance Neutron Absorption in Two-Region Problems. |
| GECINX | Abstract | P00193 H6000 00 | A Code System for Collapsing Multigroup Cross Sections in CCCC Format. |
| GENRD | Abstract | P00040 C6600 00 | Free Format Card Input Processor. |
| GENRD | Abstract | P00040 I0360 00 | Free Format Card Input Processor. |
| GERES | Abstract | P00241 I0370 00 | A Code to Produce Cross-Section Libraries for ANISN Based on Heterogeneous Fast Reactor Cell Calculations Using MC2II Data. |
| GGC-3 | Abstract | P00012 I3565 00 | Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
| GGC-3 & GGC-4 | Abstract | P00012 I3675 00 | Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
| GGC-4 | Abstract | P00012 U1108 00 | Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
| GGTC-ENEL | Abstract | P00128 I0360 00 | Code System for Producing Few-Group Neutron Cross Sections from Multigroup Data Libraries. |
| GIFT | Abstract | P00124 C0076 00 | A Combinatorial Geometry Code System with Model Testing Routines. |
| GIFT | Abstract | P00124 D0VAX 00 | A Combinatorial Geometry Code System with Model Testing Routines. |
| GIFT | Abstract | P00124 U0000 00 | A Combinatorial Geometry Code System with Model Testing Routines. |
| GIP | Abstract | P00229 IBMPC 00 | Group-Organized Cross-Section Input Program. |
| GLUCS | Abstract | P00192 D0VAX 00 | A Generalized Least-Squares Code System for Updating Cross Section Evaluations with Correlated Data Sets. |
| GNASH-FKK | Abstract | P00535 MNYCP 00 | Pre-equilibrium, Statistical Nuclear-Model Code System for Calculation Cross Sections and Emission Spectra. |
| GROUPXS | Abstract | P00246 C0740 00 | Processing of Double-Differential Cross Sections in the New ENDF-VI Format. |
| HAUSER*5 | Abstract | P00152 U0000 00 | Code System for Calculating Nuclear Cross Sections. |
| HEITLER | Abstract | P00004 I7030 00 | Cross Section Generator. |
| INGEN | Abstract | P00207 C0000 00 | A General-Purpose Mesh Generator for Finite Element Codes. |
| LAFPX-V | Abstract | D00054 C0000 01 | A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections. |
| LAFPX-V | Abstract | D00054 C0000 02 | A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections. |
| LAPHANO | Abstract | P00020 C6600 00 | PO Multigroup Photon Production Matrix and Source Vector Code for ENDF Data. |
| LAPHANO | Abstract | P00020 I0360 00 | PO Multigroup Photon Production Matrix and Source Vector Code for ENDF Data. |
| LEAP-ADDELT | Abstract | P00138 I0360 00 | Multigroup Thermal Neutron Scattering Data Generator for Hydrogen in Light Water and Deuterium in Heavy Water. |
| LIBMAK | Abstract | P00087 I0360 00 | ANISN-Type Binary Data Processing Code System. |
| LOOM-P | Abstract | P00153 F2307 00 | A Finite Element Mesh Generation Code System with On-Line Graphic Display. |
| MACK-IV | Abstract | P00132 I3691 00 | Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format. |
| MANYFILE | Abstract | P00068 I0360 00 | Utility Routine - Manipulation of Data Sets Between Various I-O Devices. |
| MARCOPOLO | Abstract | P00225 I0360 00 | Code System for Calculating the Radial and Axial Neutron Diffusion Coefficients in One-Group and Multigroup Theory. |
| MARIA SYSTEM | Abstract | P00359 D6000 00 | Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations. |
| MARS | Abstract | P00117 I0360 00 | Collection of Computer Codes for Manipulating Multigroup Cross Section Libraries in AMPX or CCCC Formats. |
| MC**2-2 | Abstract | P00350 SUN05 01 | Multigroup Cross Section Generation Code for Fast Reactor Analysis. |
| MC**2-3 | Abstract | P00577 MNYCP 00 | Multigroup Cross Section Generation Code for Fast Reactor Analysis. |
| MC**2-3 EXE | Abstract | P00577 MNYCP 01 | Multigroup Cross Section Generation Code for Fast Reactor Analysis. |
| MCB63NEA.BOLIB | Abstract | D00216 MNYCP 00 | ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code. |
| MCJEF22NEA.BOLIB | Abstract | D00203 MNYCP 01 | JEF 2.2 Cross Section Library for the MCNP Monte Carlo Code. |
| MICROX-2 | Abstract | P00374 MNYCP 02 | Code System to Create Broad-Group Cross Sections with Resonance Interference and Self-Shielding from Fine-Group and Pointwise Cross Sections. |
| MIGROS3 | Abstract | P00265 I0370 00 | A Code for the Generation of Group Constants for Reactor Calculations from Neutron Nuclear Data in KEDAK Format. |
| MINIGAL | Abstract | P00180 I3033 00 | Neutron Cross Section Processing System for Calculating Average Values from Data in the Standard United Kingdom Nuclear Data Library Format. |
| MINX | Abstract | P00105 C6600 00 | Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats. |
| MINX | Abstract | P00105 I0360 00 | Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats. |
| MIXEN | Abstract | P00318 IRISC 00 | Code System to Replace Files 4 and 6 of ENDF-6 with Files 4 and 5 of ENDF/B-IV. |
| MSM-SOURCE | Abstract | P00369 MNYCP 00 | Code System for Generation of Input Data for MCNP. |
| MUXS | Abstract | P00187 I3033 00 | Generator of Multigroup Cross Sections for Charged Particle Transport Problems. |
| NANICK | Abstract | P00120 I0360 00 | Infinitely-Diluted Multigroup Cross-Section Generator - from ENDF/B. |
| NASIF-NARES | Abstract | P00121 I0360 00 | A Code System for Computing Shielding Factors from ENDF/B Tapes. |
| NEUPAC | Abstract | P00177 FM200 00 | Neutron Unfolding Code System for Calculating Neutron Flux Spectra from Activation Data of Dosimeter Foils. |
| NJOY91.119 | Abstract | P00171 MFMWS 04 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY94.61 | Abstract | P00355 MFMWS 03 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY97.0 | Abstract | P00368 MNYCP 00 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY99.0 | Abstract | P00480 MNYCP 00 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY-UTIL-EIR | Abstract | P00296 C0825 00 | Utilities For the NJOY (6/83) Nuclear Data Processing System. |
| NPTXS | Abstract | P00090 I0360 00 | Data Generator: Neutron Point Cross Sections from ENDF/B Resolved and Unresolved Resonance Parameters. |
| NSLINK | Abstract | P00314 D0VAX 00 | NJOY SCALE LINK. |
| NUCWIZ | Abstract | P00616 PCX86 00 | NucWiz |
| NUFACE | Abstract | P00284 CYXMP 00 | An Interface Code For The Calculation of Nuclear Responses. |
| O5S | Abstract | P00014 DP010 00 | Response Function Generator--An O5R Monte Carlo Code for Calculating Pulse Height Distributions Due to Monoenergetic Neutrons Incident on Organic Scintillators. |
| O5S | Abstract | P00014 I3675 00 | Response Function Generator--An O5R Monte Carlo Code for Calculating Pulse Height Distributions Due to Monoenergetic Neutrons Incident on Organic Scintillators. |
| ORIGEN-JENDL32 | Abstract | C00703 MNYWS 00 | Isotope Generation and Depletion Code - Matrix Exponential Method. |
| PADF-2007 | Abstract | D00259 PCX86 00 | Proton Activation Data File in ENDF-6 Format. |
| PAPER 1 | Abstract | P00097 C6600 00 | Monte Carlo Calculation of Solid Angle and Self-Absorption Factors for an Inclined Cylindrical Source Viewed by a Cylindrical Detector. |
| PAPIN | Abstract | P00156 I0370 00 | A Code System to Calculate Cross Section Probability Tables, Bondarenko and Transmission Self-Shielding Factors for Fertile Isotopes in the Unresolved Resonance Region. |
| PIXSE | Abstract | P00133 I0360 00 | A Generator of Multigroup and Multipoint Cross Sections for Thermal Reactor Calculations. |
| PLASMX | Abstract | P00106 C6600 00 | A Multigroup Ionization and Charge Exchange Cross-Section Code System for Neutral Hydrogen Transport in Plasmas. |
| POLLA | Abstract | P00208 I3033 00 | A Fortran Program to Convert R-MATRIX-Type Multilevel Resonance Parameters for Fissile Nuclei into Equivalent KAPUR-PEIERLS-Type Parameters. |
| POPOP4 | Abstract | P00011 I3675 00 | Converter of Gamma-Ray Spectra to Secondary Gamma-Ray Production Cross Sections. |
| PRE-ANISN | Abstract | P00332 PC386 00 | A Preprocessing Code for ANISN and Other Radiation Transport Codes. |
| PRECO2006 | Abstract | P00226 MNYCP 02 | Exciton Model Code System for Calculating Preequilibrium and Direct Double Differential Cross Sections. |
| PREPRO2019 | Abstract | P00351 MNYCP 10 | Pre-Processing Code System for Data in ENDF/B Format. |
| PRIMEDANA-2 | Abstract | C00490 I3081 00 | Collapses Multigroup Cross Sections and Obtains Reaction Parameters by Solving Transport or Diffusion Equations. |
| PUFF-IV | Abstract | P00534 MNYCP 01 | Determination of Multigroup Covariance Matrices from ENDF/B-V Uncertainty Files. |
| RADAK | Abstract | P00122 I0360 00 | Flux Spectra Unfolding Code System - Neutron or Gamma-Ray Detectors. |
| RADHEAT-V4 | Abstract | C00300 FM380 00 | A Code System To Generate Multigroup Constants and Analyze Radiation Transport for Shielding Safety Evaluation. |
| RGENDF | Abstract | P00239 C0170 00 | Format Translation from NJOY GENDF Format to ENDF/B-V and Other Formats. |
| RICE | Abstract | P00022 I0360 00 | A Program to Calculate Primary Recoil Atom Spectra from ENDF/B Data. |
| ROLAIDS-CPM | Abstract | P00353 SUN04 00 | Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method. |
| RSYST | Abstract | C00269 I0360 00 | Integrated Modular Code System for Shielding and Reactor Physics Calculations. |
| S1CALC | Abstract | P00134 I0360 00 | A Multigroup Thermal Neutron Scattering Law Data Generator for Hydrogen and Deuterium. |
| SAMMY 8.1.0 | Abstract | P00158 MNYCP 13 | Code System for Multilevel R-Matrix Fits to Neutron and Charged-Particle Cross-Section Data Using Bayes' Equations. |
| SATURN | Abstract | P00057 I3675 00 | P1 or Transport Corrected Multigroup Neutron Cross Section Data Processor. |
| SCAT-2 | Abstract | P00294 MNYCP 03 | Code System for Calculating Total and Elastic Scattering Cross Sections Based on an Optical Model of the Spherical Nucleus. |
| SLAROM | Abstract | P00244 FM380 00 | A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors. |
| SMOG | Abstract | P00216 I3033 00 | Code System for Neutron Cross Section Evaluation (Optical Method). |
| SPECTER | Abstract | P00023 I3565 00 | Calculation of Energy Distribution of Nuclear Reaction Products. |
| SPHINX | Abstract | P00129 C7600 00 | A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code System. |
| SPHINX | Abstract | P00129 I0360 00 | A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code System. |
| STAPREF | Abstract | P00498 PC586 00 | Code System to Calculate Nuclear Reaction Cross Sections by Evaporation Model. |
| SUPERDAN-PC | Abstract | P00282 IBMPC 00 | Calculates Dancoff Factor of Spheres, Cylinders and Slabs. |
| SUPERTOG III M2 | Abstract | P00013 I3691 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-4 | Abstract | P00013 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| TALYS-1.2 | Abstract | P00548 PC586 01 | Nuclear Model Code System for Analysis and Prediction of Nuclear Reactions and Generation of Nuclear Data. |
| TDOWN-IV | Abstract | P00172 H6000 00 | A Code System to Generate Composition- and Spatially-Dependent Neutron Cross Sections for Multigroup Neutronics Analysis. |
| TECALC | Abstract | P00074 DP010 00 | Interactive Calculation of Compton Coherent and Photoelectric Mass Attenuation Coefficients for Photons (E<1 MeV), and the Mass Absorption Coefficient for Known Materials. |
| TENDL-2008-ACE | Abstract | D00243 MNYCP 00 | TALYS-Based Cross Section Library for Use with MCNP(X). |
| TENDL-2010-ACE | Abstract | D00248 MNYCP 00 | TALYS-Based Cross Section Library for Use with MCNP(X). |
| THRUSH | Abstract | P00276 CYXMP 00 | Calculates Thermal Neutron Scattering Kernel. |
| TIMS-1 | Abstract | P00163 D0780 00 | Processing Code System for Production of Group Constants of Heavy Resonant Nuclei. |
| TIMS-1 | Abstract | P00163 FM200 00 | Processing Code System for Production of Group Constants of Heavy Resonant Nuclei. |
| TRANSX 2.15 | Abstract | P00317 MFMWS 01 | Code system to produce neutron, photon, and particle transport tables for discrete-ordinates and diffusion codes from cross sections in MATXS format. |
| UNF | Abstract | P00521 PC586 00 | Code System to Calculate Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials. |
| UNIFY-ECN | Abstract | P00288 C0170 00 | A Program to Calculate Fast Neutron Data for Structural Materials. |
| UPEML 3.0 | Abstract | P00245 ALLCP 01 | A Machine-Portable CDC UPDATE Emulator. |
| URR | Abstract | P00281 D6220 00 | Calculates Resonance Neutron Cross-Section Probability Tables, Bondarenko Self-Shielding Factors and Self-Indication Ratios for Fissile and Fertile Nuclides. |
| UTXS6 | Abstract | D00211 MNYCP 00 | MCNP Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1365K. |
| XLACS-IIA | Abstract | P00182 I3033 00 | A Modified Version of XLACS-II for Processing ENDF Data into Multigroup Neutron Cross Sections in AMPX Master Library Format. |
| XSUN-2013 | Abstract | C00825 PCX86 00 | Windows interface environment for transport and sensitivity-uncertainty software TRANSX-2, PARTISN and SUSD3D |
| YUMMY | Abstract | D00221 MNYCP 00 | Multi-temperature, Neutron Cross Section Library Based on ENDF/B-V and ENDF/B-VI for use with MCNP. |