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| Packages with Keyword: NEUTRON CROSS SECTION PROCESSING |
| Package Name | Abstract | RSICC Tapelist | Title |
| 1DX | Abstract | P00096 U1108 00 | A One-Dimensional Diffusion Code System for Producing Energy Group Collapsed and Self-Shielded Cross Sections. |
| ADLER III | Abstract | P00058 I0360 00 | A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters. |
| AMARA | Abstract | P00079 I3675 00 | Nuclear Data Adjustment Using Lagrange's Multipliers Method. |
| AMPX-77 | Abstract | P00315 ALLMF 01 | Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
| AXMIX-PC | Abstract | P00297 IBMPC 00 | ANISN Cross Section Code System. |
| BREESE-II | Abstract | P00143 I3033 00 | Auxiliary Routines for Implementing the Albedo Option in the MORSE Monte Carlo Code System. |
CALENDF-2010
OECD | Abstract | P00578 PCX86 00 | Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations. |
| CODAC (2) | Abstract | P00073 I0360 00 | For TIMOC 72, Monte Carlo Three-Dimensional Neutron Transport Code's Data Generator. |
| COMBINE-PC | Abstract | P00286 IBMPC 00 | Code System to Compute Neutron Spectra and ENDF/B Version 5 Based Multigroup Neutron Constants. |
| EDITOR | Abstract | P00035 I0360 00 | Alters Mode, Copies, Merges, Punches, Edits, or Adds to ENDF/B-Formatted Data on Tapes or Cards. |
| ELAN | Abstract | P00141 ICL00 00 | Neutron Cross-Section Self-Shielding Code System. |
| ELIESE-3 | Abstract | P00003 I0370 00 | Analyses of Elastic and Inelastic Scattering Cross Sections. |
| ENBAL2 | Abstract | P00160 I0370 00 | A Program to Generate Multigroup Neutron Kerma Factors. |
| 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. |
| ERIC-2 | Abstract | P00119 I0360 00 | Calculator of Resonance Integral and Effective Capture and Fission Cross Sections for Fissile and Non-Fissile Nuclides - Thermal or Fast Reactors. |
| ETHEL | Abstract | P00217 I0360 00 | Code System for Generating Cross Sections for PSR-128/THERMOS. |
| 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. |
| FEDGROUP-R | Abstract | P00349 MNYCP 00 | 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. |
| FITOCO | Abstract | P00189 C0175 00 | Converter of Fine-Group Flux Density and Cross Section Data to Coarse Group Values. |
| FOURACES | Abstract | P00183 I0370 00 | Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries. |
| GALAXY-6 | Abstract | P00098 I0370 00 | Neutron Multigroup Cross Section Processor. |
| GAROL | Abstract | P00033 I7090 00 | Calculation of Resonance Neutron Absorption in Two-Region Problems. |
| GECINX | Abstract | P00193 H6000 00 | A Code System for Collapsing Multigroup Cross Sections in CCCC Format. |
| GERES | Abstract | P00241 I0370 00 | A Code to Produce Cross-Section Libraries for ANISN Based on Heterogeneous Fast Reactor Cell Calculations Using MC2II Data. |
| GGC-3 | Abstract | P00012 I3565 00 | Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
| GGC-3 & GGC-4 | Abstract | P00012 I3675 00 | Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
| GGC-4 | Abstract | P00012 U1108 00 | Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
| GGTC-ENEL | Abstract | P00128 I0360 00 | Code System for Producing Few-Group Neutron Cross Sections from Multigroup Data Libraries. |
| GIP | Abstract | P00229 IBMPC 00 | Group-Organized Cross-Section Input Program. |
| GLUCS | Abstract | P00192 D0VAX 00 | A Generalized Least-Squares Code System for Updating Cross Section Evaluations with Correlated Data Sets. |
| LEAP-ADDELT | Abstract | P00138 I0360 00 | Multigroup Thermal Neutron Scattering Data Generator for Hydrogen in Light Water and Deuterium in Heavy Water. |
| LIBMAK | Abstract | P00087 I0360 00 | ANISN-Type Binary Data Processing Code System. |
| LSL-M2 | Abstract | P00233 D6220 00 | Least-Squares Logarithmic Adjustment of Neutron Spectra. |
| LSL-M2 | Abstract | P00233 IBMPC 00 | Least-Squares Logarithmic Adjustment of Neutron Spectra. |
| MARCOPOLO | Abstract | P00225 I0360 00 | Code System for Calculating the Radial and Axial Neutron Diffusion Coefficients in One-Group and Multigroup Theory. |
| MICAP | Abstract | P00261 I3033 00 | A Monte Carlo Code System for Analysis of Ionization Chamber Responses. |
| MIGROS3 | Abstract | P00265 I0370 00 | A Code for the Generation of Group Constants for Reactor Calculations from Neutron Nuclear Data in KEDAK Format. |
| MINIGAL | Abstract | P00180 I3033 00 | Neutron Cross Section Processing System for Calculating Average Values from Data in the Standard United Kingdom Nuclear Data Library Format. |
| 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. |
| 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. |
| 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. |
| PIXSE | Abstract | P00133 I0360 00 | A Generator of Multigroup and Multipoint Cross Sections for Thermal Reactor Calculations. |
| PUFF-IV | Abstract | P00534 MNYCP 01 | Determination of Multigroup Covariance Matrices from ENDF/B-V Uncertainty Files. |
| 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. |
| RICE | Abstract | P00022 I0360 00 | A Program to Calculate Primary Recoil Atom Spectra from ENDF/B Data. |
| ROLAIDS-CPM | Abstract | P00353 SUN04 00 | Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method. |
| S1CALC | Abstract | P00134 I0360 00 | A Multigroup Thermal Neutron Scattering Law Data Generator for Hydrogen and Deuterium. |
| SAIPS | Abstract | P00203 E1040 00 | Information Processing System for Calculating Neutron Spectra from Measured Reaction Rates. |
| SAIPS-PC | Abstract | P00295 IBMPC 00 | Information Processing System for Calculating Neutron Spectra from Measured Reaction Rates. |
| SATURN | Abstract | P00057 I3675 00 | P1 or Transport Corrected Multigroup Neutron Cross Section Data Processor. |
| SCAT-2 | Abstract | P00294 MNYCP 03 | Code System for Calculating Total and Elastic Scattering Cross Sections Based on an Optical Model of the Spherical Nucleus. |
| SLAROM | Abstract | P00244 FM380 00 | A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors. |
| SPHINX | Abstract | P00129 C7600 00 | A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code System. |
| SPHINX | Abstract | P00129 I0360 00 | A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code System. |
| SUPERTOG-JR. | Abstract | P00115 F2307 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-JR. | Abstract | P00115 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| TDOWN-IV | Abstract | P00172 H6000 00 | A Code System to Generate Composition- and Spatially-Dependent Neutron Cross Sections for Multigroup Neutronics Analysis. |
| THERMOS-OTA | Abstract | P00107 C0173 00 | Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
| THERMOS-OTA | Abstract | P00107 C0740 00 | Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
| THERMOS-OTA | Abstract | P00107 U1108 00 | Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
| TIMS-1 | Abstract | P00163 D0780 00 | Processing Code System for Production of Group Constants of Heavy Resonant Nuclei. |
| TIMS-1 | Abstract | P00163 FM200 00 | Processing Code System for Production of Group Constants of Heavy Resonant Nuclei. |
| UKE-III | Abstract | P00015 I3691 00 | Cross Section Format Translator - UKNDL to ENDF/B. |
| URR | Abstract | P00281 D6220 00 | Calculates Resonance Neutron Cross-Section Probability Tables, Bondarenko Self-Shielding Factors and Self-Indication Ratios for Fissile and Fertile Nuclides. |
| XLACS-IIA | Abstract | P00182 I3033 00 | A Modified Version of XLACS-II for Processing ENDF Data into Multigroup Neutron Cross Sections in AMPX Master Library Format. |