| Packages with Subject: CROSS-SECTION DATA, FACTORS, AND COEFFICIENTS |
| 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. |
| ACTV-FUS/INT | Abstract | D00170 ALLCP 00 | International Library of Neutron Activation Cross-Section Data for Fusion Reactor Application. |
| 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. |
| 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. |
| BOXER | Abstract | C00766 MNYWS 00 | Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations |
| BREESE-II | Abstract | P00143 I3033 00 | Auxiliary Routines for Implementing the Albedo Option in the MORSE Monte Carlo Code System. |
| 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. |
| CARP-82 | Abstract | P00131 I3033 00 | Multigroup Albedo Data Using DOT Angular Flux Results. |
| 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. |
| 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. |
| COV-15GROUP-2006 | Abstract | D00232 MNYCP 00 | 15-Group Cross Section Covariance Matrix Library. |
| COVERX | Abstract | D00044 I0360 02 | Compilation of Multigroup Cross-Section Covariance Matrices in COVERX Format for Several Important Materials. |
| COVFILS-2 | Abstract | D00137 ALLCP 00 | Neutron Data and Covariances for Sensitivity and Uncertainty Analysis. |
| 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. |
| DANCOFF-MC | Abstract | P00509 MNYCP 00 | Code System for Monte Carlo Calculation of Dancoff Factors in Irregular Geometries. |
| DDXLIB | Abstract | D00123 FM380 01 | 125-Neutron Group Double Differential Cross Section Library. |
| DECAYREM | Abstract | D00030 I0360 02 | Radioactive Decay Spectra in EXREM Format. |
| DOSCOV | Abstract | D00090 I0360 00 | 24-Group Covariance Data. |
| 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. |
| ECPL82 | Abstract | D00106 ALLCP 00 | Evaluated Charged-Particle Data Library. |
| ELAST2 | Abstract | D00208 MNYCP 00 | Database of Cross Sections for the Elastic Scattering of Electrons and Positrons by Atoms. |
| ELIESE-3 | Abstract | P00003 I0370 00 | Analyses of Elastic and Inelastic Scattering Cross Sections. |
| 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. |
| 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-2.1 | 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. |
| FGXRRS | Abstract | D00132 C0000 00 | Few Group Cross Section Library for Research Reactor Calculations. |
| 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. |
| FLUNG | Abstract | D00086 I3033 00 | Coupled 35-Group Neutron and 21-Group Gamma Ray, P3 Cross Sections for Fusion Applications. |
| FORSS | Abstract | C00334 C0000 00 | A Sensitivity and Uncertainty Analysis Code System. |
| FORSS | Abstract | C00334 I0360 00 | A Sensitivity and Uncertainty Analysis Code System. |
| 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-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. |
| GAMTAB | Abstract | D00032 I0360 00 | Radioactive-Decay Gamma-Rays Ordered by Energy and Nuclide. |
| 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. |
| HELLO | Abstract | D00058 I0360 00 | 47 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 60 MeV. |
| HILO2K | Abstract | D00220 MNYCP 00 | Group Cross Sections for Radiation Transport |
| HPICE | Abstract | D00007 I0360 05 | Evaluated Photon Interaction Library, ENDF/B File 23 Format. |
| IEAF-2001 | Abstract | D00217 MNYCP 00 | Intermediate Energy Activation File - 2001. |
| JENDL/D-99 | Abstract | D00204 MNYCP 00 | JENDL Dosimetry File 99. |
| 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. |
| 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. |
| LUMP | Abstract | D00089 I0360 00 | Evaluated Lumped Fission Product Cross Sections for Fast Reactor Analysis--Based on ENDF/B-V Data. |
| 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. |
| MATXS175/42-JE | Abstract | D00151 D8810 00 | JEF/EFF Based VITAMIN-J 175 Neutron, 42 Photon Multigroup Data Library in MATXS Format. |
| MATXS70-JEF87 | Abstract | D00148 D8810 00 | JEF/EFF Based 70 Group Neutron Data Library in MATXS Format. |
| 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. |
| NCSP-DAT | Abstract | M00002 MNYCP 01 | Nuclear Data in Support of the Nuclear Criticality Safety Program. |
| NMTC/JAERI97 | Abstract | C00694 SUN05 00 | Monte Carlo Nucleon Meson Transport Code System. |
| NMTC/JAM | Abstract | C00717 PC586 00 | High Energy Particle Transport Code System. |
| 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. |
| ORIGEN-JENDL32 | Abstract | C00703 MNYWS 00 | Isotope Generation and Depletion Code - Matrix Exponential Method. |
| 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. |
| 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. |
| POPLIB | Abstract | D00012 I0360 03 | A Compendium of Neutron-Induced Secondary Gamma-Ray Yield and Cross Section Data. |
| 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. |
| 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. |
| SENPRO | Abstract | D00045 I3691 02 | Compilation of Multigroup Sensitivity Profiles in SENPRO Format for Fast Reactor Core and Shield Benchmarks and Thermal Reactor Benchmarks. |
| SENSIT | Abstract | C00405 C7600 00 | One-Dimensional, Multigroup Cross Section and Design Sensitivity and Uncertainty Analysis Code System - Generalized Perturbation Theory. |
| 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. |
| 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. |
| SUGGEL | Abstract | P00508 MNYWS 00 | Program Suggesting the Orbital Angular Momentum of a Neutron Resonance From the Magnitude Of Its Neutron Width. |
| SUSD | Abstract | C00501 HM150 00 | Cross Section Sensitivity and Uncertainty Analysis Including Secondary Neutron Energy and Angular Distributions. |
| SUSD | Abstract | C00501 I3090 00 | Cross Section Sensitivity and Uncertainty Analysis Including Secondary Neutron Energy and Angular Distributions. |
| SUSD3D | Abstract | C00695 MNYCP 01 | Multi-Dimensional, Discrete-Ordinates Based Cross Section Sensitivity and Uncertainty Analysis Code System. |
| SWANLAKE | Abstract | C00204 C6600 00 | Cross Section Sensitivity Analysis Code System for One-Dimensional Discrete Ordinates Calculations. |
| SWANLAKE | Abstract | C00204 I3033 00 | Cross Section Sensitivity Analysis Code System for One-Dimensional Discrete Ordinates Calculations. |
| 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). |
| THERMGAM | Abstract | D00140 ALLCP 00 | Prompt Gamma Rays from Thermal-Neutron Capture. |
| TRIGLAV | Abstract | P00495 PC586 00 | Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor. |
| 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. |
| 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. |
| 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. |
| WIMKAL-88 | Abstract | D00193 MNYCP 00 | 69 Energy Group, Neutron Cross Section Library For Thermal Reactor Calculations in WIMSD Format. |
| WIMS-ANL 4.0 | Abstract | C00698 MNYCP 00 | Deterministic Code System for Reactor Lattice Calculation. |
| 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. |
| WLUP 3.0 | Abstract | D00231 MNYCP 01 | 69- and 172- Group Cross Section Libraries for WIMS. |
| 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. |