| Packages with Keyword: NEUTRON CROSS SECTIONS |
| Package Name | Abstract | RSICC Tapelist | Title |
| ABBN-90 | Abstract | D00182 MNYCP 00 | Multigroup Constant Set for Calculation of Neutron and Photon Radiation Fields and Functionals, Including the CONSYST2 Program. |
| ACTL82 | Abstract | D00069 ALLCP 01 | Evaluated Neutron Activation Cross-Section Library. |
| ACTV-F/H | Abstract | D00155 ALLCP 00 | Neutron Activation Cross Section Library for Fusion Reactor Design. |
| ALEPH-LIB-JEFF3.1 | Abstract | D00230 MNYCP 00 | ACE Format Neutron Cross Section Library based on JEFF3.1. |
| ANSL-V | Abstract | D00154 ALLCP 01 | ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies. |
| BARC-35 | Abstract | D00124 IBMMF 00 | 35-Group Neutron Cross Sections and Resonance Self-Shielding Factors Generated in ISOTXS and BRKOXS Format from ENDF/B-IV Using MINX. |
| 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. |
| CANDULIB-AECL | Abstract | D00210 MNYCP 00 | Burnup-Dependent ORIGEN-S Cross Section Libraries for CANDU Reactor Fuel Characterization. |
| 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. |
| COVFILS-2 | Abstract | D00137 ALLCP 00 | Neutron Data and Covariances for Sensitivity and Uncertainty Analysis. |
| CRYO-S(A,B)-ACE1 | Abstract | D00253 MNYCP 00 | Scattering Law and Continuous Energy Cross Section Library of Materials at Cryogenic Temperatures. |
| DABL69 | Abstract | D00130 I0360 01 | Defense Nuclear Applications Broad-Group Library based on ENDF/B-V in ANISN Format. |
| DDXLIB | Abstract | D00123 FM380 01 | 125-Neutron Group Double Differential Cross Section Library. |
| DOSDAM77-81 | Abstract | D00081 C6400 00 | Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses. |
| DOSDAM81-82 | Abstract | D00097 C0000 00 | Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses. |
| DOSDAM84 | Abstract | D00131 IBMMF 00 | Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses. |
| 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. |
| ENDL82 | Abstract | D00103 ALLCP 00 | Neutron Library in Transmittal Format. |
| EPR | Abstract | D00037 I3691 05 | Coupled 100-Group Neutron 21-Group Gamma-ray Cross Sections for EPR Neutronics. |
| EPR MASTER | Abstract | D00052 I3691 00 | 100 Neutron Group Cross Sections in AMPX Master Library Format. |
| EURLIB-III | Abstract | D00035 I0360 01 | 100 Neutron, 20 Gamma-Ray Group Cross Section Library for Use in the European Shielding Benchmark Program. |
| 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. |
| 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. |
| 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. |
| GEAF-1 | Abstract | D00158 D8810 00 | 100 Group Cross Sections for Neutron Activation. |
| 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. |
| 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 |
| IRAN-LIB | Abstract | D00159 IBMPC 00 | A P-3 Coupled Neutron-Gamma Cross Section Library in ISOTXS For Use with ANISN/PC (CCC-514). |
| IRDF-2002 | Abstract | D00229 MNYCP 01 | The International Reactor Dosimetry File. |
| IRDF82 | Abstract | D00094 I0360 00 | The International Reactor Dosimetry File. |
| 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. |
| KEDAK3 | Abstract | D00141 I0370 00 | Evaluated Neutron Nuclear Data for Reactor Physics Calculations. |
| L26P3S34 | Abstract | D00112 IBMMF 00 | ENDL 26-Group up to P3 Library Prepared by SUPERTOG for 34 Materials. |
| 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. |
| LENDL | Abstract | D00034 I0360 02 | Livermore Evaluated Neutron and Secondary Gamma-Ray Production Cross-Section Library in ENDF/B-IV Format. |
| LIB123 | Abstract | D00153 ALLCP 00 | AMPX-II P3 123-Group Neutron Cross Section Master Interface Library. |
| 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. |
| MATXS1 | Abstract | D00114 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS10 | Abstract | D00176 ALLCP 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS11 | Abstract | D00177 ALLCP 00 | Neutron, Photon Cross Sections from ENDF/B-IV 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. |
| MCJEFF3.1NEA | Abstract | D00228 MNYCP 00 | Neutron Cross Section Library Based on JEFF3.1 for Use with MCNP. |
| MENSLIB | Abstract | D00084 I0370 00 | 60 Group, P5, Cross Sections in DTF-IV for Transport Calculations for Neutrons with Energies Up to 60 MeV. |
| MGCLIB | Abstract | D00118 FM380 00 | 137 and 26 Neutron Multigroup Cross Section Library with the Bondarenko Type Shielding Table. |
| NPCSL-81 | Abstract | D00082 I0370 00 | Point Neutron Cross Sections Generated from ENDF/B-IV with the NPTXS Modules of PSR-63/AMPX-II. |
| 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. |
| REFIT-2009 | Abstract | C00775 PCX86 00 | Multilevel Resonance Parameter Least Square Fit of Neutron Transmission, Capture, Fission & Self Indication Data. |
| SNLRML | Abstract | D00178 ALLCP 00 | Recommended Dosimetry Cross Section Compendium. |
| SUGGEL | Abstract | P00508 MNYWS 00 | Program Suggesting the Orbital Angular Momentum of a Neutron Resonance From the Magnitude Of Its Neutron Width. |
| TEMPEST-2 | Abstract | P00558 I0360 00 | Thermalization Program for Neutron Spectra and MultiGroup Cross-Sections. |
| 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). |
| TENDL-2011-ACE | Abstract | D00252 MNYCP 00 | TALYS-Based Cross Section Library for Use with MCNP(X). |
| TENDL-2012-ACE | Abstract | D00266 MNYCP 00 | TALYS-Based Cross Section Library for Use with MCNP(X). |
| TSL-ACE/2013 | Abstract | D00270 ALLCP 00 | TSL-ACE/2013 |
| UKCTRI-81 | Abstract | D00064 I0370 01 | 46-Group Neutron Cross Sections and Kerma Factors for Fusion Reactor Calculations. |
| UKNDL | Abstract | D00039 I0370 00 | United Kingdom Evaluated Neutron Cross-Section Data Library. |
| UKNDL-81 | Abstract | D00107 I3033 00 | The Aldermaston Nuclear Data Library. |
| VELM | Abstract | D00133 I0360 00 | Multigroup Cross-Section Libraries Based on ENDF/B-V Data for Sodium-Cooled Reactor Shield Analysis. |
| VITAMIN-4C | Abstract | D00053 I3691 00 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-B6 | Abstract | D00184 ALLCP 00 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-B7/BUGLE-B7 | Abstract | D00245 MNYCP 01 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| 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/COVA | Abstract | D00157 D8810 00 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-J/COVA/EFF | Abstract | D00197 ALLCP 00 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| 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. |
| WLUP 3.0 | Abstract | D00231 MNYCP 01 | 69- and 172- Group Cross Section Libraries for WIMS. |
| 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. |
| YUMMY | Abstract | D00221 MNYCP 00 | Multi-temperature, Neutron Cross Section Library Based on ENDF/B-V and ENDF/B-VI for use with MCNP. |