| Packages with Keyword: MULTIGROUP CROSS SECTIONS |
| 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. |
| ABBN-90 | Abstract | D00182 MNYCP 00 | Multigroup Constant Set for Calculation of Neutron and Photon Radiation Fields and Functionals, Including the CONSYST2 Program. |
| AIR DATA | Abstract | D00014 I0360 00 | Sample Input to ANISN for Calculation of Neutron and Secondary Gamma-Ray Transport in Air. |
| AMPX01 | Abstract | D00027 I3675 02 | Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
| AMPX-77 | Abstract | P00315 ALLMF 01 | Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
| ANSL-V | Abstract | D00154 ALLCP 01 | ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies. |
| AXMIX-PC | Abstract | P00297 IBMPC 00 | ANISN Cross Section Code System. |
| BABEL | Abstract | D00104 I3033 00 | Multi-Purpose Neutron and Gamma-Ray Cross Section Library for Fast Reactor Shielding Design. |
| BP | Abstract | D00008 I0360 00 | Data for Selected Shielding Benchmark Problems Specified in ORNL-RSIC-25, Shielding Benchmark Problems. |
| BUGLE-80 | Abstract | D00075 IBMPC 03 | Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. |
| BUGLE-80 | Abstract | D00075 PC386 01 | Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. |
| BUGLE-93 | Abstract | D00175 ALLCP 01 | Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. |
| BUGLE-96 | Abstract | D00185 ALLCP 00 | Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. |
| CASK | Abstract | D00023 I3691 04 | 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis. |
| CASK-81 | Abstract | D00023 I0370 05 | 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis. |
| CASK-81 | Abstract | D00023 IBMPC 06 | 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis. |
| CLAW-IV | Abstract | D00036 I0360 02 | Coupled 30 Neutron, 12 Gamma-Ray Group Cross Sections Based on ENDF/B-IV for Radiation Transport Calculations. |
| CLAW-IV | Abstract | D00036 I3033 03 | Coupled 30 Neutron, 12 Gamma-Ray Group Cross Sections Based on ENDF/B-IV for Radiation Transport Calculations. |
| CLEAR | Abstract | D00042 I3691 00 | 126 Neutron, 36 Gamma-Ray Cross Sections in AMPX and CCCC Interface Formats for LMFBR Neutronics Calculations. |
| COBB | Abstract | D00016 I3675 01 | 123-Group Neutron Cross Section Data Generated from ENDF/B-II Data for Use in the XSDRN Discrete Ordinates Spectral Averaging Code. |
| 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. |
| CTR DATA | Abstract | D00028 I3675 01 | 73-Group P3 Coupled Neutron and Gamma-Ray Cross Sections for Fusion Reactor Calculations. |
| DABL69 | Abstract | D00130 I0360 01 | Defense Nuclear Applications Broad-Group Library based on ENDF/B-V in ANISN Format. |
| 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. |
| DPL-400 GEDT1 | Abstract | D00031 I0360 08 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| DPL-401 NEDT | Abstract | D00031 I0360 09 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| DPL-402A/GPDT1 | Abstract | D00031 I0360 10 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| DPL-402B/GPDT1 | Abstract | D00031 I0360 11 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| E3LWR | Abstract | D00098 C0000 00 | 45 Neutron, 16 Gamma-Ray and 15 Neutron, 5 Gamma-Ray Group LWR Cross Section Libraries Derived from EURLIB-III using the AGRUKO Optimized Collapsing Scheme. |
| ENBAL2 | Abstract | P00160 I0370 00 | A Program to Generate Multigroup Neutron Kerma Factors. |
| ENDL82 | Abstract | D00103 ALLCP 00 | Neutron Library in Transmittal Format. |
| ENTOSAN | Abstract | P00188 C0175 00 | Code System for Calculating Fine-Group Dosimetry Cross Section Values from ENDF/B Data. |
| ENTOSAN | Abstract | P00188 D8810 00 | Code System for Calculating Fine-Group Dosimetry Cross Section Values from ENDF/B Data. |
| EPR | Abstract | D00037 I3691 05 | Coupled 100-Group Neutron 21-Group Gamma-ray Cross Sections for EPR Neutronics. |
| EURLIB-III | Abstract | D00035 I0360 01 | 100 Neutron, 20 Gamma-Ray Group Cross Section Library for Use in the European Shielding Benchmark Program. |
| FCXSEC | Abstract | D00085 PC386 01 | 22 Neutron, 21 Gamma-Ray Group Cross Section Libraries in ANISN Format for Nuclear Fuel Cycle Shielding Calculations. |
| FDMXPC | Abstract | P00322 IPCAT 00 | Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format. |
| 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. |
| FENDL-2.1 | Abstract | D00222 MNYCP 00 | Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications. |
| 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. |
| 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. |
| FLUNG | Abstract | D00086 I3033 00 | Coupled 35-Group Neutron and 21-Group Gamma Ray, P3 Cross Sections for Fusion Applications. |
| FORSEN | Abstract | P00170 I0360 00 | A Multigroup Processing Code for Use with Sensitivity Profiles to Assess the Effect of Cross Section Changes. |
| 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. |
| GAMLIB | Abstract | D00006 I0360 00 | 99-Group Neutron Cross Sections for Use in the GAM Portion of the GGC Multigroup Cross Section Code. |
| GARG | Abstract | D00073 C0000 00 | 27-Group Neutron Cross Sections in Discrete Ordinates Format Generated with FIGERO (PSR-149) from ENDF-B Data. |
| GARLIB | Abstract | D00013 I3565 01 | Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations. |
| GARLIB | Abstract | D00013 I7090 00 | Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations. |
| GAROL | Abstract | P00033 I7090 00 | Calculation of Resonance Neutron Absorption in Two-Region Problems. |
| GEAF-1 | Abstract | D00158 D8810 00 | 100 Group Cross Sections for Neutron Activation. |
| GECINX | Abstract | P00193 H6000 00 | A Code System for Collapsing Multigroup Cross Sections in CCCC Format. |
| 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. |
| GICX40 | Abstract | D00092 ALLCP 00 | Coupled 42-Neutron, 21-Gamma-Ray Group Cross Sections for 40 Elements in Group Independent Form for Fusion Reactor Calculations. |
| GIP | Abstract | P00229 IBMPC 00 | Group-Organized Cross-Section Input Program. |
| GROUPXS | Abstract | P00246 C0740 00 | Processing of Double-Differential Cross Sections in the New ENDF-VI Format. |
| HELLO | Abstract | D00058 I0360 00 | 47 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 60 MeV. |
| HILO | Abstract | D00087 I0370 00 | Group Cross Sections for Radiation Transport |
| HILO2K | Abstract | D00220 MNYCP 00 | Group Cross Sections for Radiation Transport |
| HILO86 | Abstract | D00119 I0360 00 | Group Cross Sections for Radiation Transport |
| HILO86 | Abstract | D00119 PC386 01 | Group Cross Sections for Radiation Transport |
| HILO86R | Abstract | D00187 ALLCP 00 | Group Cross Sections for Radiation Transport |
| IEAF-2001 | Abstract | D00217 MNYCP 00 | Intermediate Energy Activation File - 2001. |
| JFS | Abstract | D00111 I3033 00 | 70 Group Neutron Fast Reactor Cross Section Set and 25 Group Neutron Fast Reactor Cross Section Set. |
| JFS3J2 | Abstract | D00108 FM200 00 | 70 Group Neutron Fast Reactor Cross Section Set Based on JENDL-2B. |
| JIMCOF | Abstract | D00078 F2307 00 | Multigroup Constants fFle Based on ENDF/B IV. |
| LAHIMACK | Abstract | D00128 I0360 00 | A Multigroup Library of Neutron and Gamma Cross Sections and Response Functions in the Energy Range up to 800 MeV. |
| LEAP-ADDELT | Abstract | P00138 I0360 00 | Multigroup Thermal Neutron Scattering Data Generator for Hydrogen in Light Water and Deuterium in Heavy Water. |
| LIB123 | Abstract | D00153 ALLCP 00 | AMPX-II P3 123-Group Neutron Cross Section Master Interface Library. |
| LIBMAK | Abstract | P00087 I0360 00 | ANISN-Type Binary Data Processing Code System. |
| MACK-IV | Abstract | P00132 I3691 00 | Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format. |
| MACKLIB | Abstract | D00029 I3675 00 | A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV. |
| MACKLIB-IV-82 | Abstract | D00060 I0360 01 | A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV. |
| MARS | Abstract | P00117 I0360 00 | Collection of Computer Codes for Manipulating Multigroup Cross Section Libraries in AMPX or CCCC Formats. |
| MATJEFF31.BOLIB | Abstract | D00242 MNYCP 00 | Fine-Group Cross Section Library Based on JEFF3.1 for Nuclear Fission Applications. |
| MATXS1 | Abstract | D00114 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS175/42-JE | Abstract | D00151 D8810 00 | JEF/EFF Based VITAMIN-J 175 Neutron, 42 Photon Multigroup Data Library in MATXS Format. |
| MATXS5A | Abstract | D00115 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS6A | Abstract | D00116 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS70-JEF87 | Abstract | D00148 D8810 00 | JEF/EFF Based 70 Group Neutron Data Library in MATXS Format. |
| MATXS7A | Abstract | D00117 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MENSLIB | Abstract | D00084 I0370 00 | 60 Group, P5, Cross Sections in DTF-IV for Transport Calculations for Neutrons with Energies Up to 60 MeV. |
| 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. |
| MONTUK-80 | Abstract | D00072 ALLCP 01 | UKCTR III Transmutation and Activation Data, 100-Group Neutron Activation Cross-Section Data for Fusion Reactor Structure and Coolant Materials. |
| NAB | Abstract | D00018 I0360 00 | 100-Group, P3, Neutron Cross Section Data for Sodium and Aluminum. |
| NANICK | Abstract | P00120 I0360 00 | Infinitely-Diluted Multigroup Cross-Section Generator - from ENDF/B. |
| 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. |
| NJOY-UTIL-EIR | Abstract | P00296 C0825 00 | Utilities For the NJOY (6/83) Nuclear Data Processing System. |
| NOX | Abstract | D00017 I0360 00 | 199-Group, P5, Coupled Neutron and Secondary Gamma-Ray Cross Section Data for Nitrogen and Oxygen. |
| NSLINK | Abstract | P00314 D0VAX 00 | NJOY SCALE LINK. |
| ORYX-E | Abstract | D00038 I0360 00 | ORIGEN Yields and Cross Sections Nuclear Transmutation and Decay Data from ENDF/B-IV. |
| ORYX-E | Abstract | D00038 I0360 01 | ORIGEN Yields and Cross Sections Nuclear Transmutation and Decay Data from ENDF/B-IV. |
| 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. |
| POPOP4 | Abstract | P00011 I3675 00 | Converter of Gamma-Ray Spectra to Secondary Gamma-Ray Production Cross Sections. |
| PUCOR | Abstract | D00067 I3691 00 | 84 Group Neutron Cross Sections for Uranium-Plutonium Cycle LWR and PWR Models in AMPX Master Library Format. |
| PVC | Abstract | D00048 I3691 00 | 36 Group, P5, Photon Interaction Cross Sections for 38 Materials in ANISN Format. |
| RITTS | Abstract | D00011 I0360 00 | 121-Group Coupled Neutron and Gamma-Ray Cross-Section Data for Transport Codes. |
| ROLAIDS-CPM | Abstract | P00353 SUN04 00 | Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method. |
| S1CALC | Abstract | P00134 I0360 00 | A Multigroup Thermal Neutron Scattering Law Data Generator for Hydrogen and Deuterium. |
| SATURN | Abstract | P00057 I3675 00 | P1 or Transport Corrected Multigroup Neutron Cross Section Data Processor. |
| SCAMPI | Abstract | P00352 MNYWS 01 | Collection of Codes for Manipulating Multigroup Cross Section Libraries in AMPX Format. |
| SHAMSI | Abstract | D00135 I3033 00 | 48 Group Cross-Section Library for Fusion Nucleonics Analysis. |
| 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. |
| SUPERTOG-JR. | Abstract | P00115 F2307 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-JR. | Abstract | P00115 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| TDOWN-IV | Abstract | P00172 H6000 00 | A Code System to Generate Composition- and Spatially-Dependent Neutron Cross Sections for Multigroup Neutronics Analysis. |
| THERMOS-OTA | Abstract | P00107 C0173 00 | Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
| THERMOS-OTA | Abstract | P00107 C0740 00 | Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
| THERMOS-OTA | Abstract | P00107 U1108 00 | Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
| 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. |
| TRANSX-CTR | Abstract | P00206 CY000 00 | Interfaces MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis. |
| UKCTRI-81 | Abstract | D00064 I0370 01 | 46-Group Neutron Cross Sections and Kerma Factors for Fusion Reactor Calculations. |
| VITAMIN-C | Abstract | D00041 I0360 02 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-E | Abstract | D00113 I3033 02 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-J/KERMA | Abstract | D00150 I3090 00 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITENEA-J | Abstract | D00238 MNYCP 00 | AMPX 175-n,42-g Multigroup X-section Library for Nuclear Fusion Applications. |
| VITJEFF31.BOLIB | Abstract | D00235 MNYCP 00 | A JEFF-3.1 Multigr Coupled (199n + 42gamma) X-Section Lib. in AMPX Fmt for Nuclear Fission Applications. |
| WIMSLIB-IJS0 | Abstract | D00147 D8810 00 | Extended Version of the WIMS 69-group Library. |
| WIMSLIB-IJS1 | Abstract | D00147 D8810 01 | Extended Version of the WIMS 69-group Library. |
| WIMSLIB-JEF87 | Abstract | D00095 D0VAX 00 | Extended Version of the WIMS 69-group Library. |
| W-M-NRSM | Abstract | D00026 U1108 00 | WANL-MSFC Nuclear Rocket Shielding Methods Data Generator (GAMLEG-W, APPROPOS, NAGS, and SATURN) and Multigroup Neutron and Gamma-ray Cross Section Libraries 1-6. |
| 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. |