Online Catalog
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Note: RESTRICTIONS APPLY TO SOME PACKAGES -
810 -- US DOE 10CFR810 Jurisdiction
FEDC -- US Government Agencies and Their Contractors Only
OECD -- Restricted/See Abstract
USSO -- US Distribution Only
USUNV -- US Universities Only
Packages starting with M
Package NameAbstractRSICC TapelistTitle
MACK-IVAbstractP00132 I3691 00Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format.
MACKLIBAbstractD00029 I3675 00100 Group Neutron Kerma Factors and Reaction Cross Sections Generated by MACK from Data in ENDF Format.
MACKLIB-IV-82AbstractD00060 I0360 01A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV.
MADONNAAbstractC00425 I0370 00Two-dimensional Neutron Streaming Coupled Removal-Diffusion-Albedo-Transport Code System.
MAEROSAbstractP00466 C7600 00Code System for Multicomponent Aerosol Time Evolution.
MAGIKAbstractC00359 I0360 00A Monte Carlo Code System for Computing Induced Residual Activation Dose Rates.
MAGNAAbstractC00158 C3600 00Multi-Source Gamma-Ray Kernel Integration Code System.
MAINTAINAbstractP00067 I0360 00Code System for Use in Maintaining and Revising Card Image Files on Tape.
MANYFILEAbstractP00068 I0360 00Utility Routine - Manipulation of Data Sets Between Various I-O Devices.
MAPAbstractC00150 I3675 00Kernel Integration Code System in Complex Geometry with Special Application to Surface Sources Determined by Discrete Ordinates Calculations.
MARCH2AbstractP00473 CDCMF 00Code System to Model LWR Meltdown Accident Response.
MARCOPOLOAbstractP00225 I0360 00Code System for Calculating the Radial and Axial Neutron Diffusion Coefficients in One-Group and Multigroup Theory.
MARC-PNAbstractC00311 D8810 00A Neutron Diffusion Code System with Spherical Harmonics Option.
MARD 4.16AbstractP00448 IBMPC 00Models And Results Database System.
MARIA SYSTEMAbstractP00359 D6000 00Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations.
MARINRADAbstractC00503 C1785 00Code System Model for Assessing the Consequences of Release of Radioactive Material into the Oceans.
MARLOWE 15BAbstractP00137 MNYCP 08Computer Simulation of Atomic Collisions in Crystalline Solids (Version 15).
MARMERAbstractC00579 D8350 00A Flexible Point-Kernel Shielding Code System.
MARMERAbstractC00579 PC486 00Flexible Point-Kernal Shielding Code System.
MARSAbstractP00117 I0360 00Collection of Computer Codes for Manipulating Multigroup Cross Section Libraries in AMPX or CCCC Formats.
MARTHAAbstractP00232 I0360 00Monte Carlo Response Function Calculation for Sodium Iodide Photon Detectors.
MARVIKEN-JIT
OECD
AbstractD00269 MNYCP 00Marviken Full Scale Jet Impingement Tests Experiments.
MASSAbstractD00025 I0360 01Atomic Mass Evaluation.
MATADORAbstractC00689 CDCMF 00Radionuclide Behavior in Containments.
MATEXPAbstractP00059 I0360 00Matrix Exponential Method Applied to Systems of Ordinary Differential Equations.
MATJEFF31.BOLIBAbstractD00242 MNYCP 00Fine-Group Cross Section Library Based on JEFF3.1 for Nuclear Fission Applications.
MATXS1AbstractD00114 C0000 0030-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-IV in MATXS Format.
MATXS10AbstractD00176 ALLCP 0030-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-VI in MATXS Format.
MATXS11AbstractD00177 ALLCP 0080-Group Neutron, 24-Group Photon Cross Sections from ENDF/B-VI in MATXS Format.
MATXS175/42-JEAbstractD00151 D8810 00JEF/EFF Based VITAMIN-J 175 Neutron, 42 Photon Multigroup Data Library in MATXS Format.
MATXS5AAbstractD00115 C0000 0030-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-V in MATSX Format.
MATXS6AAbstractD00116 C0000 0080-Group Neutron, 24-Group Photon Fast-Reactor Cross Section from ENDF/B-V in MATXS Format.
MATXS70-JEF87AbstractD00148 D8810 00JEF/EFF Based 70 Group Neutron Data Library in MATXS Format.
MATXS7AAbstractD00117 C0000 0069-Group Thermal-Reactor Neutron Cross Section Data from ENDF/B-V in MATXS Format.
MATXSLIBJ33AbstractD00258 MNYCP 01JENDL-3.3 Based, 175 N-42 Photon Groups (VITAMIN-J) MATXS Library for Discrete Ordinates Multi-Group.
MATXUFAbstractP00130 I0360 00On-Line Derivative Method, Spectrum Unfolding Code System for NE-213 Liquid Fast Scintillation Proton Recoil Data.
MAVRACAbstractC00023 I7090 00Model Astronaut and Vehicle Radiation Analysis Code.
MAX-XTREMEAbstractP00001 C0000 00Generalized Several-Constraint LaGrange Multiplier.
MAZE IIAbstractP00041 U1108 00Spectral Unfolding Code.
MAZE-1AbstractP00041 C6600 00Spectral Unfolding Code.
MC**2-2AbstractP00350 SUN05 01Code System for Calculating Fast Neutron Spectra and Multigroup Cross-sections from ENDF/B Data (November 2000 Version).
MC**2-3AbstractP00577 MNYCP 00Multigroup Cross Section Generation Code for Fast Reactor Analysis.
MCART
USUNV
AbstractC00809 PCX86 00Solve the Time-Dependent Neutron Transport Equation.
MCB1CAbstractC00719 MNYWS 00Monte-Carlo Continuous Energy Burnup Code System.
MCB63NEA.BOLIBAbstractD00216 MNYCP 00ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code.
MCFLAREAbstractC00093 I7090 00Monte Carlo Code to Simulate Solar Flare Events and Estimate Probable Doses Encountered on Interplanetary Missions.
MCJEF22NEA.BOLIBAbstractD00203 MNYCP 01JEF 2.2 Cross Section Library for the MCNP Monte Carlo Code.
MCJEFF3.1NEAAbstractD00228 MNYCP 00Neutron Cross Section Library Based on JEFF3.1 for Use with MCNP.
MCNP6.1/MCNP5/MCNPX
810
AbstractC00810 MNYCP 00Monte Carlo N–Particle Transport Code System Including MCNP6.1, MCNP5-1.60, MCNPX-2.7.0 and Data Libraries.
MCNP6.1/MCNP5/MCNPX-EXE
810
AbstractC00810 MNYCP 01Monte Carlo N–Particle Transport Code System Including MCNP6.1, MCNP5-1.60, MCNPX-2.7.0 and Data Libraries.
MCNPDATA
810
AbstractD00200 ALLCP 03Standard Neutron, Photon, and Electron Data Libraries for MCNP4C or MCNP-PoliMi.
MCNP-DSP
810
AbstractC00699 MNYCP 00Monte Carlo N-Particle Transport Code System with Digital Signal Processing based on MCNP4A.
MCNPX-POLIMI
810
AbstractC00791 MNYCP 00Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities.
MCNPXS
810
AbstractD00189 ALLCP 00Standard Neutron, Photon, and Electron Data Libraries for MCNP4B or MCNP-DSP.
MCRACAbstractC00562 IBMPC 00Multiple Cycle Reactor Analysis Code.
MCRTOFAbstractC00435 FM200 00Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region.
MCRTOFAbstractC00435 I0360 00Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region.
MCUNED
810
AbstractC00804 PCX86 00MCNPX Extension for Using Light Ion Evaluated Nuclear Data Library.
MCVIEWAbstractP00202 FM780 00View Factor Calculation for Three-Dimensional Geometries.
MECC-7AbstractC00156 I0360 00Medium-Energy Intranuclear Cascade Code System.
MEDUSA-IBAbstractC00505 HM200 00One-Dimensional Lagrangian Code for Plasma Hydrodynamic Analysis of a Fusion Pellet Driven by Ion Beams.
MEDUSA-PIJAbstractC00349 F2307 00One-Dimensional Laser Fusion Analyzer (Including Neutron Heating Effect) Collision Probability Method.
MENDL-2PAbstractD00207 MNYCP 00Proton Reaction Data Library for Nuclear Activation (Medium Energy Nuclear Data Library.)
MENSLIBAbstractD00084 I0370 0060 Group, P5, Cross Sections in DTF-IV for Transport Calculations for Neutrons with Energies Up to 60 MeV.
MERCURE 4-82AbstractC00142 I3033 00Three-Dimensional Code System for Integrating Multigroup Line-of-Sight Attenuation Kernels by Monte Carlo Techniques.
MESAAbstractP00223 I3033 00Non-Linear Least Squares Spectral Analysis.
MESODIF-IIAbstractC00498 D0780 00A Variable Trajectory Plume Segment Model to Assess Ground-Level Air Concentrations and Depositions of Routine Effluent Releases from Nuclear Power Facilities.
MESOIAbstractC00497 D0780 00Interactive Mesoscale Lagrangian Puff Dispersion Model with Deposition and Decay. See CCC-677/MESORAD.
MESORAD 1.4AbstractC00677 D0VAX 00Code System for Emergency Response Dose Assessment.
MESYSTAbstractC00706 MNYWS 00Code System to Simulate 3D Tracer Dispersion in Atmosphere.
METDAbstractP00197 DGMV1 00Computer Code Systems for Use with Meteorological Data.
METDAbstractP00197 I3033 00Computer Code Systems for Use with Meteorological Data.
MEVDPAbstractC00157 C6600 00Primary Radiation Transport Code System - Complex Geometry - Computerized Anatomical Model Man.
MGA8AbstractP00542 MNYCP 00Code System to Determine Pu Isotope Abundances from Multichannel Analyzer Gamma Spectra.
MGCLIBAbstractD00118 FM380 00137 and 26 Neutron Multigroup Cross Section Library with the Bondarenko Type Shielding Table.
MICAPAbstractP00261 I3033 00A Monte Carlo Code System for Analysis of Ionization Chamber Responses.
MICROX-2AbstractP00374 MNYCP 02Code System to Create Broad-Group Cross Sections with Resonance Interference and Self-Shielding from Fine-Group and Pointwise Cross Sections.
MIGROS3AbstractP00265 I0370 00A Code for the Generation of Group Constants for Reactor Calculations from Neutron Nuclear Data in KEDAK Format.
MILDOSAbstractC00398 C0000 00Calculation of Radiation Doses from Uranium Recovery Operations.
MILDOS-AREAAbstractC00608 IBMPC 00Calculation of Radiation Dose from Uranium Recovery Operations for Large-Area Sources.
MINETAbstractP00490 CY000 00Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis.
MINIGALAbstractP00180 I3033 00Neutron Cross Section Processing System for Calculating Average Values from Data in the Standard United Kingdom Nuclear Data Library Format.
MINTEQAbstractP00494 DVX11 00Code System to Model Aqueous Geochemical Equilibria.
MINXAbstractP00105 C6600 00Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats.
MINXAbstractP00105 I0360 00Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats.
MISSIONARYAbstractP00114 I0360 00ENDF/B to NDL Data Format Converter.
MIXENAbstractP00318 IRISC 00Code System to Replace Files 4 and 6 of ENDF-6 with Files 4 and 5 of ENDF/B-IV.
MKENO-DARAbstractC00513 FM380 00Direct Angular Representation Monte Carlo Code for Criticality Safety Analysis
MMCRAbstractC00441 FM200 00Multigroup Monte Carlo Neutron and Photon Transport Code.
MMRWAbstractM00018 MNYCP 00Canadian and Early British Energy Reports on Nuclear Reactor Theory (1940-1946).
MOCAAbstractC00590 IPCAT 00Monte Carlo Criticality Code System for Hexagonal Geometries.
MOCUPAbstractP00365 DALPU 00MCNP/ORIGEN Coupling Utility Programs.
MODELAbstractC00329 I3033 00Models of Trapped Proton and Electron Environments for Solar Maximum and Minimum.
MOMENT IAbstractC00188 U1108 00Moments Method Neutron Transport Code System.
MOMGEM-MOMDISAbstractC00085 I7090 00Moments Method Reconstruction of Scattered Gamma-Ray Distributions.
MONK 6.3
FEDC
AbstractC00393 I3033 00A General Purpose Monte Carlo Neutronics Code System.
MONTEBURNS 2.0AbstractP00455 MNYCP 02An Automated, Multi-Step Monte Carlo Burnup Code System.
MONTUK-80AbstractD00072 ALLCP 01UKCTR III Transmutation and Activation Data, 100-Group Neutron Activation Cross-Section Data for Fusion Reactor Structure and Coolant Materials.
MORECAAbstractP00411 PC386 00Computer Code System for Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup.
MORNAbstractP00062 I0360 00Calculation of the Response of Sodium Iodide Crystals to Gamma Rays.
MORSE-ALBAbstractC00394 FM200 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System, Albedo Version. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-ANSI STD.AbstractC00127 I3675 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-BAbstractC00368 I0370 00General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CAbstractC00431 C7600 00Monte Carlo Multigroup Neutron Code System for the Solution of Criticality Problems. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAbstractC00203 C0000 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAbstractC00203 CY000 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAbstractC00203 D0VAX 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAbstractC00203 I0360 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAbstractC00203 U0000 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAAbstractC00474 ALLCP 03A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Array Geometry Capability, Version 2.
MORSEC-SP2AbstractP00142 H6000 00A Multigroup Cross Section Module for the MORSE Monte Carlo Computer Code System.
MORSE-CVAbstractC00535 HM280 00Multigroup Neutron and Gamma-Ray Monte Carlo Transport Code with Covariance Calculation. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-EAbstractC00258 I0360 00Special Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-EMPAbstractC00588 IBMPC 00General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Array Geometry Capability. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-HAbstractC00471 I3081 00A Revised Version of the MORSE Monte Carlo Radiation Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-LAbstractC00261 C6600 00Multigroup Neutron and Gamma-Ray Transport Code System for the Solution of Penetration Problems. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-SGCAbstractC00277 C7600 00A Super Grouped Cross Section Version of the MORSE Code System. We recommend either C00474/ALLCP/02 MORSE-CGA, or C00545/IRISC/01 SCALE 4.2.
MOSRA-LIGHTAbstractP00505 MNYWS 00High-Speed Three-Dimensional Nodal Diffusion Code System.
MOXY-MOD32AbstractP00385 I0360 00BWR Core Heat Transfer Code System.
MRIPP 1.0
810
AbstractC00655 PC386 00Magnetic Resonance Image Phantom Code System to Calibrate in vivo Measurement Systems.
MRSPAKAbstractP00212 DVX11 00A Code System To Generate a Text File Containing Combinatorial Geometry Data Corresponding to PADL2 Geometry.
MSM-SOURCEAbstractP00369 MNYCP 00Code System for Generation of Input Data for MCNP.
MTR_PC 2.6AbstractC00674 PC386 00Modular Code System for Neutronics, Thermalhydraulics and Shielding Calculations.
MULTI-KENO2AbstractC00492 FM380 00A Monte Carlo Code System for Criticality Safety Analysis.
MUP2AbstractP00289 I3090 00A Program to Calculate Fast Neutron Data for Medium-Heavy Nuclei.
MUREAbstractC00764 MNYWS 00MCNP Utility for Reactor Evolution.
MURLIAbstractC00378 DP011 00Integral Transport Theory Code System for Thermal Reactor Lattice Cell Calculation.
MUSCATAbstractC00281 I0360 00Calculation of Neutron Currents in Spherical and Cylindrical Cavities by Means of View Factors.
MUSPALBAbstractC00171 ICL00 00Albedo Calculation of Multigroup Spectra of Neutrons Transmitted Through Multilayer Slab Shielding.
MUXSAbstractP00187 I3033 00Generator of Multigroup Cross Sections for Charged Particle Transport Problems.
MVP-GMVP IIAbstractC00739 MNYCP 00General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods.
MYRAAbstractC00056 C0000 00Calculation of Shipping Costs and Cask Designs for Irradiated Fuel Elements.
MYRAAbstractC00056 I7090 00Calculation of Shipping Costs and Cask Designs for Irradiated Fuel Elements.
The Radiation Safety Information Computational Center (RSICC) is a Department of Energy Specialized Information Analysis Center (SIAC) authorized to collect, analyze, maintain, and distribute computer software and data sets in the areas of radiation transport and safety. RSICC resides in the Reactor and Nuclear Systems Division (RNSD) at Oak Ridge National Laboratory.