| Data Library Collection (nuclear data libraries, etc.) |
| 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. |
| ACTIV87 | Abstract | D00169 ALLCP 00 | Fast Neutron Activation Cross Section File. |
| 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. |
| ACTV-FUS/INT | Abstract | D00170 ALLCP 00 | International Library of Neutron Activation Cross-Section Data for Fusion Reactor Application. |
| ADS-LIB/V2.0 | Abstract | D00250 MNYCP 00 | Test Library for Accelerator Driven Systems V2.0 |
| AGDATA | Abstract | D00127 I0360 00 | Two Agricultural Production Data Libraries (AGDATC and AGDATG) for Dose and Risk Assessment Models. |
| AIR DATA | Abstract | D00014 I0360 00 | Sample Input to ANISN for Calculation of Neutron and Secondary Gamma-Ray Transport in Air. |
| AIRFEWG | Abstract | D00049 I0360 00 | Results of ANISN Multigroup Calculations of Gamma-Ray, Neutron, and Secondary Gamma-Ray Transport in Infinite Homogeneous Air Using DLC-31/(DPL-1/FEWG1) Cross Sections. |
| ALBEDO-DATA | Abstract | D00224 MNYCP 00 | KSU Neutron Albedo Data. |
| ALEPH-LIB-JEFF3.1 | Abstract | D00230 MNYCP 00 | ACE Format Neutron Cross Section Library based on JEFF3.1. |
| AMPX01 | Abstract | D00027 I3675 02 | Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
| ANS643 | Abstract | D00129 IBMPC 02 | Geometric Progression Gamma-Ray Buildup Factor Coefficients. |
| ANSL-V | Abstract | D00154 ALLCP 01 | ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies. |
| BABEL | Abstract | D00104 I3033 00 | Multi-Purpose Neutron and Gamma-Ray Cross Section Library for Fast Reactor Shielding Design. |
| 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. |
| BOREHOLE-EB6.8-MG | Abstract | D00268 MNYCP 00 | Multi-Group Cross-Section Library for Deterministic and Monte Carlo Codes. |
| BP | Abstract | D00008 I0360 00 | Data for Selected Shielding Benchmark Problems Specified in ORNL-RSIC-25, Shielding Benchmark Problems. |
| BUGENDF70.BOLIB | Abstract | D00262 PCX86 00 | ENDF/B-VII.0 Broad-Group Coupled Cross Section Library for LWR Shielding & Pressure Vessel Dosimetry Applications. |
| BUGJEFF311.BOLIB | Abstract | D00254 MNYCP 01 | JEFF-3.1.1 Broad-Group Coupled Cross Section Library For LWR Shielding & Pressure Vessel Dosimetry Applications. |
| 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. |
| CAD | Abstract | D00059 I0360 00 | 51 Neutron, 25 Gamma-Ray Group ALBEDO DATA Generated with DOT for Various Materials. |
| CANDULIB-AECL | Abstract | D00210 MNYCP 00 | Burnup-Dependent ORIGEN-S Cross Section Libraries for CANDU Reactor Fuel Characterization. |
| CASK | Abstract | D00023 I3691 04 | 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. |
| CASK-81 | Abstract | D00023 I0370 05 | 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. |
| CLES | Abstract | D00233 MNYCP 00 | Cross Section Library of Moderator Materials for Low-Energy Neutron Sources. |
| 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. |
| COG SUPPLEMENTAL LIBRARIES | Abstract | D00271 MNYCP 00 | COG LibMaker – Data Conversion Utility |
| COV-15GROUP-2006 | Abstract | D00232 MNYCP 00 | 15-Group Cross Section Covariance Matrix Library. |
| COVERV | Abstract | D00077 I0360 01 | Compilation of Multigroup Cross-Section Covariance Matrices in COVERX Format for Several Important Materials. |
| COVERX | Abstract | D00044 I0360 02 | Compilation of Multigroup Cross-Section Covariance Matrices in COVERX Format for Several Important Materials. |
| COVFILS | Abstract | D00091 I0360 00 | Neutron Data and Covariances for Sensitivity and Uncertainty Analysis. |
| 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. |
| 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. |
| DDXLIB | Abstract | D00123 FM380 01 | 125-Neutron Group Double Differential Cross Section Library. |
| DECAYREM | Abstract | D00030 I0360 02 | Radioactive Decay Spectra in EXREM Format. |
| DECDC 1.0 | Abstract | D00213 MNYCP 00 | Nucear Decay Data Files for Radiation Dosimetry Calculations. |
| DOSCOV | Abstract | D00090 I0360 00 | 24-Group Covariance Data. |
| 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. |
| DOSDAT II-81 | Abstract | D00079 I0370 00 | Dose-Rate Conversion Factors for External Exposure to Photons and Electrons. |
| DOSDAT-DOE | Abstract | D00144 ALLMF 00 | Dose-Rate Conversion Factors for External Exposure to Photons and Electrons. |
| DOSDAT-DOE | Abstract | D00144 IBMPC 01 | Dose-Rate Conversion Factors for External Exposure to Photons and Electrons. |
| 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. |
| DRALIST | Abstract | D00080 ALLCP 00 | Radioactive Decay Data for Application to Radiation Dosimetry and Radiological Assessments. |
| 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. |
| EACRP-D2O-LATTICES | Abstract | D00264 MNYCP 00 | Compilation of Reactor Physics Measurements in HWRs Lattices. |
| ECPL82 | Abstract | D00106 ALLCP 00 | Evaluated Charged-Particle Data Library. |
EDSFI
USSO | Abstract | D00215 PC486 00 | Electrical Distribution System Functional Inspection Data Base. |
| ELAST2 | Abstract | D00208 MNYCP 00 | Database of Cross Sections for the Elastic Scattering of Electrons and Positrons by Atoms. |
| ELECSPEC | Abstract | D00100 DP010 00 | Electron Spectra from Decay of Fission Products. |
| ENDL82 | Abstract | D00103 ALLCP 00 | Neutron Library in Transmittal Format. |
| ENDLIB-97 | Abstract | D00179 MNYCP 01 | LLNL Libraries of Atomic Data, Electron Data, and Photon Data in Evaluated Nuclear Data Library (ENDL) Type Format. |
| ENSL82-CDRL82 | Abstract | D00110 ALLCP 00 | Evaluated Nuclear Structure Libraries. |
| EPICS2014 | Abstract | D00272 MNYCP 00 | Electron Photon Interaction Cross Sections |
| EPICS2017 | Abstract | D00272 MNYCP 01 | Electron Photon Interaction Cross Sections |
| 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. |
| ESG | Abstract | D00065 I0360 00 | 56-Group Cross Section Library Based on VITAMIN-C Generated by Using SPHINX and XSDRNPM to Collapse 171 Groups. |
| 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. |
| 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. |
| 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. |
| FIREDATA | Abstract | D00125 PC486 00 | Nuclear Power Plant Fire Data Base for Personal Computers. |
| FIS-PROD | Abstract | D00152 ALLCP 00 | Chinese Evaluated Fission Product Yield Library in ENDF/B-V Format. |
| 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. |
| FLUKA05-PRE-LIB | Abstract | D00260 PCX86 00 | FLUKA05 Multi-Group, Multi-Purpose Nuclear Data Library, Neutrons, Photons, Charged Particles. |
| FLUNG | Abstract | D00086 I3033 00 | Coupled 35-Group Neutron and 21-Group Gamma Ray, P3 Cross Sections for Fusion Applications. |
| FPDL | Abstract | D00066 I0360 00 | Fission Product Yields, Gamma Ray and Beta Spectra in ENDF-III Format for 235U, 238U, 239Pu, 232Th, and 233U. |
| 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. |
| GAMDAT-78 | Abstract | D00083 I0370 00 | Library of Gamma-Ray Decay Data for 2055 Radionuclides. |
| GAMLIB | Abstract | D00006 I0360 00 | 99-Group Neutron Cross Sections for Use in the GAM Portion of the GGC Multigroup Cross Section Code. |
| GAMMON | Abstract | D00071 ALLCP 00 | Gamma-Ray Moments Method Code System. |
| GAMTAB | Abstract | D00032 I0360 00 | Radioactive-Decay Gamma-Rays Ordered by Energy and Nuclide. |
| GAMTOT78 | Abstract | D00109 CY00I 00 | Compilation of Radioactive Decay and Capture Gamma Rays. |
| 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. |
| 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. |
| GROUP STRUCTURE | Abstract | D00156 ALLCP 00 | Standard Energy Group Structures Of Cross Section Libraries For Reactor Shielding, Reactor Cell Fusion Neutronics Applications: VITAMIN-J, ECC0-33, ECC0-2000. |
| GROUPSTRUCTURES | Abstract | D00274 MNYCP 00 | GROUPSTRUCTURES, VITAMIN-J, XMAS, ECCO-33, ECCO2000 Standard Group Structures |
| HALLMARK | Abstract | D00005 I0360 00 | Discrete Ordinates and Monte Carlo Results of Neutron and Secondary Gamma-Ray Transport in Air-Over-Ground Geometry. |
| HATCHES-19 | Abstract | D00206 PC586 02 | Thermodynamic Database for Radiochemical Modelling. |
| 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 |
| HPICE | Abstract | D00007 I0360 05 | Evaluated Photon Interaction Library, ENDF/B File 23 Format. |
| HPPOS 1.5 | Abstract | D00173 IBMPC 00 | Health Physics Position Database. |
| HPPOS V2 | Abstract | D00173 IBMPC 01 | Health Physics Position Database. |
| HUGO | Abstract | D00099 I3033 00 | Photon Interaction Data in ENDF/B Format. |
| HUGO VI | Abstract | D00146 I3033 00 | Photon Interaction Data in ENDF/B Format. |
| IEAF-2001 | Abstract | D00217 MNYCP 00 | Intermediate Energy Activation File - 2001. |
| 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. |
| IRDF-90 | Abstract | D00161 ALLCP 01 | The International Reactor Dosimetry File. |
| I-R-MAN | Abstract | D00050 ALLCP 00 | Photon Interaction Data on ICRP Reference Man. |
| IRPHE-VENUS-RECYCLE | Abstract | D00263 MNYCP 00 | Plutonium Recycling Physics Project Critical Experiments. |
| JENDL/D-99 | Abstract | D00204 MNYCP 00 | JENDL Dosimetry File 99. |
| JENDL-1 | Abstract | D00070 ALLCP 00 | Japanese Evaluated Neutron Cross Section Data in ENDF/B-IV Format. |
| JENDL-2 | Abstract | D00122 FM380 00 | Japanese Evaluated Neutron Cross Section Data in ENDF/B-IV Format. |
| 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. |
| KAOS/LIB-V | Abstract | D00160 CY000 00 | A Library of Nuclear Response Functions Generated by KAOS-V Code From ENDF/B-V and Other Data Files. |
| KDDK | Abstract | D00061 I0360 00 | Measured Results of Delayed Beta- and Gamma-Ray Spectra due to Thermal-Neutron Fission of U-235. |
| KEDAK3 | Abstract | D00141 I0370 00 | Evaluated Neutron Nuclear Data for Reactor Physics Calculations. |
| KERMAL | Abstract | D00142 ALLCP 00 | Neutron and Gamma-Ray Kerma Factors Based on LLNL Nuclear Data Files. |
| KX-RAY | Abstract | D00021 I0360 00 | Evaluated X-ray Cross Section Library. |
| L26P3S34 | Abstract | D00112 IBMMF 00 | ENDL 26-Group up to P3 Library Prepared by SUPERTOG for 34 Materials. |
| LA100 | Abstract | D00168 ALLCP 00 | Evaluated Nuclear Data Library for Transport Calculations Involving Incident Neutrons and Protons of Energy Up to 100 MeV. |
| 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. |
| 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. |
LAS CRUCES
USSO | Abstract | D00194 ALLCP 00 | Las Cruces Trench Site Database, Vadose Model. |
| LENDL | Abstract | D00034 I0360 02 | Livermore Evaluated Neutron and Secondary Gamma-Ray Production Cross-Section Library in ENDF/B-IV Format. |
| LENDL V | Abstract | D00120 I0360 00 | Lawrence Livermore National Laboratory Evaluated Nuclear Data Library in ENDF-V Format. |
| LEP | Abstract | D00001 I0360 02 | Cascade and Evaporation Particle Results from Low-Energy Intranuclear Cascade Calculations. |
| LIB123 | Abstract | D00153 ALLCP 00 | AMPX-II P3 123-Group Neutron Cross Section Master Interface Library. |
| LUMP | Abstract | D00089 I0360 00 | Evaluated Lumped Fission Product Cross Sections for Fast Reactor Analysis--Based on ENDF/B-V Data. |
| 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. |
| MASS | Abstract | D00025 I0360 01 | Atomic Mass Evaluation. |
| 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. |
| 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. |
| 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. |
| MATXSLIBJ33 | Abstract | D00258 MNYCP 01 | JENDL-3.3 Based, 175 Neutron-42 Photon Groups (VITAMIN-J) MATXS Library for Discrete Ordinates Multi-Group Transport Codes. |
| 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. |
| MCJEFF3.1NEA | Abstract | D00228 MNYCP 00 | Neutron Cross Section Library Based on JEFF3.1 for Use with MCNP. |
| MENDL-2P | Abstract | D00207 MNYCP 00 | Proton Reaction Data Library for Nuclear Activation (Medium Energy Nuclear Data Library.) |
| 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. |
| 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. |
| NEACRP-H2O-LATTICES | Abstract | D00265 MNYCP 00 | Compilation of Reactor Physics Measurements in LWRs Lattices. |
| NOX | Abstract | D00017 I0360 00 | 199-Group, P5, Coupled Neutron and Secondary Gamma-Ray Cross Section Data for Nitrogen and Oxygen. |
| NPCSL-81 | Abstract | D00082 I0370 00 | Point Neutron Cross Sections Generated from ENDF/B-IV with the NPTXS Modules of PSR-63/AMPX-II. |
| NUCDECAY | Abstract | D00172 PC386 01 | Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP and MIRD. |
| NUCDECAYCALC | Abstract | D00202 PC586 00 | Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP. |
| ORESUND | Abstract | D00267 MNYCP 00 | Nordic Mesoscale Dispersion Experiments over Land-Water-Land. |
| ORLIBJ32 | Abstract | D00255 MNYCP 00 | ORIGEN2 Libraries Based on JENDL-3.2. |
| 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. |
| PADF-2007 | Abstract | D00259 PCX86 00 | Proton Activation Data File in ENDF-6 Format. |
| PEFPYD | Abstract | D00096 ALLMF 02 | Aggregate Fission-Product Decay Data Based on ENDF/B-IV and -V. |
| PGAA-IAEA | Abstract | D00234 MNYCP 00 | Databsae for Prompt Gamma-Ray Neutron Activation Analysis. |
| PHOBIA | Abstract | D00236 PCX86 00 | Photon buildup factors to account for angular incidence on shield walls. |
| PHOTX | Abstract | D00136 D0VAX 01 | Photon Interaction Cross Section Library. |
| PHOTX | Abstract | D00136 IBMPC 00 | Photon Interaction Cross Section Library. |
| PIXE2010 | Abstract | D00246 MNYCP 00 | Proton/alpha Ionization (K, L, M shell), Tabulated Cross Section Library. |
| PNESD | Abstract | D00166 PC386 00 | Proton Nucleus Elastic Scattering Data. |
| POINT2015 | Abstract | D00273 MNYCP 00 | A Temperature-Dependent Linearly Interpolable, Tabulated Cross Section Library Based on ENDF/B |
| POPLIB | Abstract | D00012 I0360 03 | A Compendium of Neutron-Induced Secondary Gamma-Ray Yield and Cross Section Data. |
| PR-EDB | Abstract | D00196 IBMPC 03 | Power Reactor Embrittlement Data Base. |
| PUCOR | Abstract | D00067 I3691 00 | 84 Group Neutron Cross Sections for Uranium-Plutonium Cycle LWR and PWR Models in AMPX Master Library Format. |
| PUDK | Abstract | D00074 I0360 00 | Measured Results of Delayed Beta- and Gamma-Ray Spectra Due to Thermal-Neutron Fission of Pu239 and Pu241. |
| PVC | Abstract | D00048 I3691 00 | 36 Group, P5, Photon Interaction Cross Sections for 38 Materials in ANISN Format. |
| PVE | Abstract | D00126 I3033 00 | 38 Group, P8, Photon Interaction Cross Sections in ANISN Format from VITAMIN-E. |
| PWR-AXBUPRO-GKN | Abstract | D00209 MNYCP 00 | Measured Axial Burnup Profiles for NeckarWesthiem PWR Reactors. |
| PWR-AXBUPRO-SNL | Abstract | D00201 MNYCP 00 | Axial Burnup Profile Database for Pressurized Water Reactors. |
| RECOIL | Abstract | D00055 I3033 01 | Multigroup Primary Recoil Spectra, Displacement Rates and Gas-Production Rates for Radiation Damage Studies. |
| RITTS | Abstract | D00011 I0360 00 | 121-Group Coupled Neutron and Gamma-Ray Cross-Section Data for Transport Codes. |
| SAIL | Abstract | D00057 I0360 00 | 23 Neutron, 17 Gamma-Ray Group ALBEDO DATA for Concrete and Steel, Based on DOT 1-1/2-D Calculations using DLC-31/FEWG1 Data. |
| SAILOR | Abstract | D00076 PC386 01 | Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. |
| SENPRO | Abstract | D00045 I3691 02 | Compilation of Multigroup Sensitivity Profiles in SENPRO Format for Fast Reactor Core and Shield Benchmarks and Thermal Reactor Benchmarks. |
| SHAMSI | Abstract | D00135 I3033 00 | 48 Group Cross-Section Library for Fusion Nucleonics Analysis. |
| SIGMA-A | Abstract | D00139 ALLMF 00 | Photon Interaction and Absorption Cross Sections. |
| SIGMA-A | Abstract | D00139 IBMPC 00 | Photon Interaction and Absorption Cross Sections. |
| SINBAD-2022 | Abstract | D00237 MNYCP 06 | Shielding Integral Benchmark Archive and Database. |
| SKYDATA-KSU | Abstract | D00188 IBMPC 00 | Parameters for Approximate Neutron and Gamma-Ray Skyshine Response Functions and Ground Correction Factors. |
| SKYPORT | Abstract | D00093 IBMPC 00 | Skyshine Importance Functions for Neutrons and Gamma Rays. |
| SNLRML | Abstract | D00178 ALLCP 00 | Recommended Dosimetry Cross Section Compendium. |
| STORM-ISRAEL | Abstract | D00015 I0360 01 | Evaluated Photon Interaction Library, ENDF/B File 23 Format. |
| TDF | Abstract | D00162 ALLCP 00 | Thermonuclear Data File. |
| 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). |
| THERMGAM | Abstract | D00140 ALLCP 00 | Prompt Gamma Rays from Thermal-Neutron Capture. |
| TPASGAM 85 | Abstract | D00088 ALLCP 04 | Radioactive Decay Library of Gamma-Ray Energies, Branching Ratios, and Cross Sections. |
| TRANSMIT | Abstract | D00020 I0360 00 | Experimental Neutron Transmission Data Used to Test Total Cross Sections. |
| TR-EDB | Abstract | D00198 IBMPC 00 | Test Reactor Embrittlement Data Base. |
| 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. |
| UKFY2 | Abstract | D00171 IBMPC 00 | UK Fission Product Yield Library. |
| UKNDL | Abstract | D00039 I0370 00 | United Kingdom Evaluated Neutron Cross-Section Data Library. |
| UKNDL-81 | Abstract | D00107 I3033 00 | The Aldermaston Nuclear Data Library. |
| UNGER | Abstract | D00164 PC386 00 | Effective Dose Equivalent for Specific Radionuclides. |
| UTXS6 | Abstract | D00211 MNYCP 00 | MCNP Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1365K. |
| VELM | Abstract | D00133 I0360 00 | Multigroup Cross-Section Libraries Based on ENDF/B-V Data for Sodium-Cooled Reactor Shield Analysis. |
| VIP-MAN | Abstract | D00256 MNYCP 00 | Computational Phantom. |
| 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. |
| 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. |
| VITENDF70.BOLIB | Abstract | D00261 PCX86 00 | ENDF/B-VII.0 Multi-Group Coupled (199n +42gamma) Cross Section Library in AMPX Format for Nuclear Fission Applications. |
| VITENEA-E | Abstract | D00240 MNYCP 00 | AMPX 175-n,42-g Multigroup X-section Library for Nuclear Fusion Applications. |
| VITENEA-J | Abstract | D00238 MNYCP 00 | AMPX 175-n,42-g Multigroup X-section Library for Nuclear Fusion Applications. |
| VITJEF22.BOLIB | Abstract | D00241 MNYCP 00 | JEF-2.2 Multigroup Coupled (199n + 42?) Cross-Section Library in AMPX Format for Nuclear Fission 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. |
| VITJEFF311.BOLIB | Abstract | D00257 MNYCP 01 | JEFF-3.1.1 Multi-Group Coupled (199n + 42gamma) X-Section Library in AMPX Format for Nuclear Fission Applications. |
| WIMKAL-88 | Abstract | D00193 MNYCP 00 | 69 Energy Group, Neutron Cross Section Library For Thermal Reactor Calculations in WIMSD Format. |
| 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. |
| XCOM | Abstract | D00174 IBMPC 00 | Photon Cross Sections on a Personal Computer. |
| XG-IAEA | Abstract | D00163 IBMPC 00 | X-ray and Gamma-ray Standards For Detector Calibration. |
| YUMMY | Abstract | D00221 MNYCP 00 | Multi-temperature, Neutron Cross Section Library Based on ENDF/B-V and ENDF/B-VI for use with MCNP. |
| ZZ-PWR-MSLB | Abstract | D00275 MNYCP 00 | ZZ PWR-MSLB, PWR Main Steam-Line Break Benchmarks, Coupled Neutronics Thermal-Hydraulics |