| Package Name |
Abstract |
RSICC Tapelist |
Title |
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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. |
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 |
126-Group Coupled Neutron and Gamma-Ray Transport Cross-Section Data Generated by AMPX. |
ANS643
|
Abstract
|
D00129 IBMPC 02 |
ANS-6.4.3 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. |
BUGLE-80
|
Abstract
|
D00075 PC386 01 |
Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96. |
BUGLE-80
|
Abstract
|
D00075 IBMPC 02 |
Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96. |
BUGLE-80
|
Abstract
|
D00075 IBMPC 03 |
Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96. |
BUGLE-96
|
Abstract
|
D00185 ALLCP 00 |
Coupled 47 Neutron, 20 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 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. |
COVERV
|
Abstract
|
D00077 I0360 01 |
Compilation of Multigroup Cross-section Covariance Matrices in COVERX Format for Several Important Materials (Generated from ENDF/B-V Data using PSR-093/PUFF2). |
COVERX
|
Abstract
|
D00044 I0360 02 |
Compilation of Multigroup Cross-Section Covariance Matrices in COVERX Format for Several Important Materials. |
COVFILS
|
Abstract
|
D00091 I0360 00 |
A 30-Group Covariance Library Based on ENDF/B-V. |
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. |
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 |
620 Group, SAND-II Formatted, Neutron Cross Sections Based on ENDF/B-IV and Other Sources 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. |
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. |
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. |
FIPDOR
|
Abstract
|
D00068 I3691 00 |
126 Neutron Group Fission Product Cross Sections. |
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. |
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. |
FSXLIB-J3
|
Abstract
|
D00165 ALLCP 00 |
MCNP continuous energy neutron cross section library based on JENDL-3. See DLC-190/FSX96 based on JENDL3.2. |
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. |
GAMMON
|
Abstract
|
D00071 ALLCP 00 |
Activation Library for Fusion Reaction Application and Other Design Studies. |
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 I7090 00 |
Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations. |
GARLIB
|
Abstract
|
D00013 I3565 01 |
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. |
HALLMARK
|
Abstract
|
D00005 I0360 00 |
Discrete Ordinates and Monte Carlo Results of Neutron and Secondary Gamma-Ray Transport in Air-Over-Ground Geometry. |
HATCHES-12
|
Abstract
|
D00206 PC486 00 |
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 |
66 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 400 MeV. |
HILO2K
|
Abstract
|
D00220 MNYCP 00 |
Coupled 83 Neutron, 22 Photon Group Cross Sections for Neutron Energies Up to 2 GeV. |
HILO86
|
Abstract
|
D00119 I0360 00 |
66 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV. |
HILO86
|
Abstract
|
D00119 PC386 01 |
66 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV. |
HILO86R
|
Abstract
|
D00187 ALLCP 00 |
66 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV. |
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 Positions (HPPOS) Data Base Based on Current 10 CFR 20. |
HUGO
|
Abstract
|
D00099 I3033 00 |
Photon Interaction Data in ENDF/B-V Format. |
HUGO VI
|
Abstract
|
D00146 I3033 00 |
Photon Interaction Data in ENDF/B-VI Format. PHOTB6 in DLC-179/ENDLIB-97 is an updated version of these data. |
I-R-MAN
|
Abstract
|
D00050 ALLCP 00 |
Photon Interaction Data on ICRP Reference Man. |
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. |
IRDF-90
|
Abstract
|
D00161 ALLCP 01 |
The International Reactor Dosimetry File. |
IRDF82
|
Abstract
|
D00094 I0360 00 |
International Reactor Dosimetry Data. |
JENDL-1
|
Abstract
|
D00070 ALLCP 00 |
Japanese Evaluated Nuclear Data Library. |
JENDL-2
|
Abstract
|
D00122 FM380 00 |
Japanese Evaluated Neutron Cross Section Data in ENDF/B-IV Format. |
JENDL/D-99
|
Abstract
|
D00204 MNYCP 00 |
JENDL Dosimetry File 99. |
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 |
100 Group Neutron Kerma Factors and Reaction Cross Sections Generated by MACK from Data in ENDF Format. |
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 |
30-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-IV in MATXS Format. |
MATXS10
|
Abstract
|
D00176 ALLCP 00 |
30-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-VI in MATXS Format. |
MATXS11
|
Abstract
|
D00177 ALLCP 00 |
80-Group Neutron, 24-Group Photon Cross Sections from ENDF/B-VI 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 |
30-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-V in MATSX Format. |
MATXS6A
|
Abstract
|
D00116 C0000 00 |
80-Group Neutron, 24-Group Photon Fast-Reactor Cross Section from ENDF/B-V 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 |
69-Group Thermal-Reactor Neutron Cross Section Data from ENDF/B-V 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. |
MCNPDATA
|
Abstract
|
D00200 ALLCP 03 |
Standard Neutron, Photon, and Electron Data Libraries for MCNP4C or MCNP-PoliMi. |
MCNPXS
|
Abstract
|
D00189 ALLCP 00 |
Standard Neutron, Photon, and Electron Data Libraries for MCNP4B or MCNP-DSP. |
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. |
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. See newer version in RASCAL (CCC-553). |
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. |
PEFPYD
|
Abstract
|
D00096 ALLMF 02 |
Aggregate Fission-Product Decay Data Based on ENDF/B-IV and -V. |
PHOTX
|
Abstract
|
D00136 IBMPC 00 |
Photon Interaction Cross Section Library. |
PHOTX
|
Abstract
|
D00136 D0VAX 01 |
Photon Interaction Cross Section Library. |
PNESD
|
Abstract
|
D00166 PC386 00 |
Proton Nucleus Elastic Scattering Data. |
POINT2009
|
Abstract
|
D00239 MNYCP 00 |
A Temperature-Dependent Linearly Interpolable, Tabulated Cross Section Library Based on ENDF/B-VII.0 |
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, Version 3. |
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. |
RADDECAY 4.02
|
Abstract
|
D00134 IBMPC 03 |
Radioactive Decay Data for Radiological Assessments. |
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 I3033 00 |
Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96. |
SAILOR
|
Abstract
|
D00076 PC386 01 |
Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96. |
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 2009.02
|
Abstract
|
D00237 MNYCP 00 |
Shielding Integral Benchmark Archive and Database, Version February 2009. |
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. |
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. |
TR-EDB
|
Abstract
|
D00198 IBMPC 00 |
Test Reactor Embrittlement Data Base, Version 1. |
TRANSMIT
|
Abstract
|
D00020 I0360 00 |
Experimental Neutron Transmission Data Used to Test Total Cross Sections. |
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, Version 2. |
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. |
VITAMIN-4C
|
Abstract
|
D00053 I3691 00 |
171 Neutron Group Cross Sections and Bondarenko Factors in CCCC Interface Formats for Fusion and LMFBR Neutronics. |
VITAMIN-B6
|
Abstract
|
D00184 ALLCP 00 |
A Fine-Group Cross Section Library Based on ENDF/B-VI Release 3 for Radiation Transport Applications. |
VITAMIN-C
|
Abstract
|
D00041 I0360 02 |
171 Neutron, 36 Gamma-Ray Group Cross Sections in AMPX and CCCC Interface Formats for Fusion and LMFBR Neutronics. |
VITAMIN-E
|
Abstract
|
D00113 I3033 02 |
174n, 38g Cross-Section Library in AMPX Format. |
VITAMIN-J/COVA
|
Abstract
|
D00157 D8810 00 |
Neutron Cross-Section Covariance Data in Multigroup Form. |
VITAMIN-J/COVA/EFF
|
Abstract
|
D00197 ALLCP 00 |
Neutron Cross-Section Covariance Data in Multigroup Form. |
VITAMIN-J/KERMA
|
Abstract
|
D00150 I3090 00 |
VITAMIN-J 175-Neutron and 38-Photon Kerma And Gas Production Cross Sections. |
VITENEA-J
|
Abstract
|
D00238 MNYCP 00 |
AMPX 175-n,42-g Multigroup X-section Library for Nuclear Fusion Applications. |
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. |
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 |
JEF-1 Based 69 Group Neutron Data Library. |
WLUP 3.0
|
Abstract
|
D00231 MNYCP 00 |
69- and 172- Group Cross Section Libraries for WIMS. |
XCOM
|
Abstract
|
D00174 IBMPC 00 |
Photon Cross Sections on a Personal Computer, Versions 1.2 and 1.3. |
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. |
