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Click on Abstract to read the package abstract.
Click on RSICC Tapelist to view list of files distributed with package.
Note: RESTRICTIONS APPLY TO SOME PACKAGES -
810 -- US DOE 10CFR810 Jurisdiction
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| Packages with Keyword: ENDF FORMAT |
| Package Name | Abstract | RSICC Tapelist | Title |
| ADENA | Abstract | P00190 C0000 00 | Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra. |
| ADENA | Abstract | P00190 I3033 00 | Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra. |
| ADLER III | Abstract | P00058 I0360 00 | A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters. |
| AMPX-77 | Abstract | P00315 ALLMF 01 | Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
CALENDF-2010
OECD | Abstract | P00578 PCX86 00 | Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations. |
| CHENDF 7.02 | Abstract | P00333 MNYCP 05 | Codes for Handling ENDF/B-V and ENDF/B-VI Data. |
| CODAC (2) | Abstract | P00073 I0360 00 | For TIMOC 72, Monte Carlo Three-Dimensional Neutron Transport Code's Data Generator. |
| COMPLOT | Abstract | P00259 IBMMF 00 | Convert EXFOR Format Data to Computation Format and Plot Comparisons of EXFOR and ENDF/B Evaluated Data (Version 86-1). |
| CRECTJ5 | Abstract | P00250 D0780 00 | A Computer Program for Compilation of Evaluated Nuclear Data in ENDF/B Format. |
| EDITOR | Abstract | P00035 I0360 00 | Alters Mode, Copies, Merges, Punches, Edits, or Adds to ENDF/B-Formatted Data on Tapes or Cards. |
| ENDLIB-97 | Abstract | D00179 MNYCP 01 | LLNL Libraries of Atomic Data, Electron Data, and Photon Data in Evaluated Nuclear Data Library (ENDL) Type Format. |
| ENDVER/GUI | Abstract | P00572 PCX86 00 | The ENDF File Verification Support Package. |
| 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. |
| EVALPLOT | Abstract | P00211 I3081 00 | A Program to Plot Data in the Evaluated Nuclear Data File/Version B Format. |
| F5TAB | Abstract | P00221 D0780 00 | Code System for Converting Energy Distribution Cross Section Data to Tabulated Data. |
| 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. |
| FIGERO | Abstract | P00149 C0000 00 | Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations. |
| FORSEN | Abstract | P00170 I0360 00 | A Multigroup Processing Code for Use with Sensitivity Profiles to Assess the Effect of Cross Section Changes. |
| GECINX | Abstract | P00193 H6000 00 | A Code System for Collapsing Multigroup Cross Sections in CCCC Format. |
| GLUCS | Abstract | P00192 D0VAX 00 | A Generalized Least-Squares Code System for Updating Cross Section Evaluations with Correlated Data Sets. |
| GROUPXS | Abstract | P00246 C0740 00 | Processing of Double-Differential Cross Sections in the New ENDF-VI Format. |
| MACK-IV | Abstract | P00132 I3691 00 | Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format. |
| MENDL-2P | Abstract | D00207 MNYCP 00 | Proton Reaction Data Library for Nuclear Activation (Medium Energy Nuclear Data Library.) |
| MICROX-2 | Abstract | P00374 MNYCP 02 | Code System to Create Broad-Group Cross Sections with Resonance Interference and Self-Shielding from Fine-Group and Pointwise Cross Sections. |
| 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. |
| MISSIONARY | Abstract | P00114 I0360 00 | ENDF/B to NDL Data Format Converter. |
| MIXEN | Abstract | P00318 IRISC 00 | Code System to Replace Files 4 and 6 of ENDF-6 with Files 4 and 5 of ENDF/B-IV. |
| NANICK | Abstract | P00120 I0360 00 | Infinitely-Diluted Multigroup Cross-Section Generator - from ENDF/B. |
| NASIF-NARES | Abstract | P00121 I0360 00 | A Code System for Computing Shielding Factors from ENDF/B Tapes. |
| 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. |
| NJOY97.0 | Abstract | P00368 MNYCP 00 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY99.0 | Abstract | P00480 MNYCP 00 | 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. |
| NPTXS | Abstract | P00090 I0360 00 | Data Generator: Neutron Point Cross Sections from ENDF/B Resolved and Unresolved Resonance Parameters. |
| NSLINK | Abstract | P00314 D0VAX 00 | NJOY SCALE LINK. |
| ORPLOT-PC | Abstract | P00328 PC386 00 | Plotting Package for Data Evaluation Intercomparison. |
| 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. |
| RESENDD | Abstract | P00215 C0740 00 | A Code System for Reconstruction of Resonance Cross Sections from Evaluated Nuclear Data in ENDF/B Format. |
| RESENDD | Abstract | P00215 D0780 00 | A Code System for Reconstruction of Resonance Cross Sections from Evaluated Nuclear Data in ENDF/B Format. |
| REX2-87 | Abstract | P00290 D8810 00 | A Code For Calculating Self-Shielded Multigroup Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data Files. |
| SAMMY 8.1.0 | Abstract | P00158 MNYCP 13 | Code System for Multilevel R-Matrix Fits to Neutron and Charged-Particle Cross-Section Data Using Bayes' Equations. |
| STAY'SL | Abstract | P00113 DP010 00 | Least Squares Dosimetry Unfolding Code System. |
| STAYSL PNNL | Abstract | P00589 PCX86 00 | STAYSL PNNL Suite of Software Tools. |
| SUPERTOG III M2 | Abstract | P00013 I3691 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-4 | Abstract | P00013 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-LTT | Abstract | P00228 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| UNF | Abstract | P00521 PC586 00 | Code System to Calculate Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials. |
| VSOP94 | Abstract | C00670 MNYWS 00 | Computer Code System for Reactor Physics and Fuel Cycle Simulation. |
| WINDOWS II | Abstract | P00161 I0370 00 | A Program for the Analysis of Spectral Data Foil Activation Measurements. |