| Packages starting with M |
| Package Name | Abstract | RSICC Tapelist | Title |
| MACK-IV | Abstract | P00132 I3691 00 | Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format. |
| MACKLIB | Abstract | D00029 I3675 00 | A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV. |
| MACKLIB-IV-82 | Abstract | D00060 I0360 01 | A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV. |
| MADONNA | Abstract | C00425 I0370 00 | Two-dimensional Neutron Streaming Coupled Removal-Diffusion-Albedo-Transport Code System. |
| MAEROS | Abstract | P00466 C7600 00 | Code System for Multicomponent Aerosol Time Evolution. |
| MAGIK | Abstract | C00359 I0360 00 | A Monte Carlo Code System for Computing Induced Residual Activation Dose Rates. |
| MAGNA | Abstract | C00158 C3600 00 | Multi-Source Gamma-Ray Kernel Integration Code System. |
| MAINTAIN | Abstract | P00067 I0360 00 | Code System for Use in Maintaining and Revising Card Image Files on Tape. |
| MANYFILE | Abstract | P00068 I0360 00 | Utility Routine - Manipulation of Data Sets Between Various I-O Devices. |
| MAP | Abstract | C00150 I3675 00 | Kernel Integration Code System in Complex Geometry with Special Application to Surface Sources Determined by Discrete Ordinates Calculations. |
| MARCH2 | Abstract | P00473 CDCMF 00 | Code System to Model LWR Meltdown Accident Response. |
| MARCOPOLO | Abstract | P00225 I0360 00 | Code System for Calculating the Radial and Axial Neutron Diffusion Coefficients in One-Group and Multigroup Theory. |
| MARC-PN | Abstract | C00311 D8810 00 | A Neutron Diffusion Code System with Spherical Harmonics Option. |
| MARD 4.16 | Abstract | P00448 IBMPC 00 | Models And Results Database System. |
| MARIA SYSTEM | Abstract | P00359 D6000 00 | Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations. |
| MARINRAD | Abstract | C00503 C1785 00 | Code System Model for Assessing the Consequences of Release of Radioactive Material into the Oceans. |
| MARLOWE 15B | Abstract | P00137 MNYCP 08 | Computer Simulation of Atomic Collisions in Crystalline Solids. |
| MARMER | Abstract | C00579 D8350 00 | A Flexible Point-Kernel Shielding Code System. |
| MARMER | Abstract | C00579 PC486 00 | A Flexible Point-Kernel Shielding Code System. |
| MARS | Abstract | P00117 I0360 00 | Collection of Computer Codes for Manipulating Multigroup Cross Section Libraries in AMPX or CCCC Formats. |
| MASS | Abstract | D00025 I0360 01 | Atomic Mass Evaluation. |
| MATADOR | Abstract | C00689 CDCMF 00 | Radionuclide Behavior in Containments. |
| MATEXP | Abstract | P00059 I0360 00 | Matrix Exponential Method Applied to Systems of Ordinary Differential Equations. |
| MATJEFF31.BOLIB | Abstract | D00242 MNYCP 00 | Fine-Group Cross Section Library Based on JEFF3.1 for Nuclear Fission Applications. |
| MATXS1 | Abstract | D00114 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS10 | Abstract | D00176 ALLCP 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS11 | Abstract | D00177 ALLCP 00 | Neutron, Photon Cross Sections from ENDF/B-IV 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 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS6A | Abstract | D00116 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV 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 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXSLIBJ33 | Abstract | D00258 MNYCP 01 | JENDL-3.3 Based, 175 Neutron-42 Photon Groups (VITAMIN-J) MATXS Library for Discrete Ordinates Multi-Group Transport Codes. |
| MATXUF | Abstract | P00130 I0360 00 | On-Line Derivative Method, Spectrum Unfolding Code System for NE-213 Liquid Fast Scintillation Proton Recoil Data. |
| MAVRAC | Abstract | C00023 I7090 00 | Model Astronaut and Vehicle Radiation Analysis Code. |
| MAX-XTREME | Abstract | P00001 C0000 00 | Generalized Several-Constraint LaGrange Multiplier. |
| MAZE II | Abstract | P00041 U1108 00 | Spectral Unfolding Code. |
| MAZE-1 | Abstract | P00041 C6600 00 | Spectral Unfolding Code. |
| MC**2-2 | Abstract | P00350 SUN05 01 | Multigroup Cross Section Generation Code for Fast Reactor Analysis. |
| MC**2-3 | Abstract | P00577 MNYCP 00 | Multigroup Cross Section Generation Code for Fast Reactor Analysis. |
| MC**2-3 EXE | Abstract | P00577 MNYCP 01 | Multigroup Cross Section Generation Code for Fast Reactor Analysis. |
MCART
USUNV | Abstract | C00809 PCX86 00 | Solve the Time-Dependent Neutron Transport Equation. |
| MCB1C | Abstract | C00719 MNYWS 00 | Monte-Carlo Continuous Energy Burnup Code System. |
| MCB63NEA.BOLIB | Abstract | D00216 MNYCP 00 | ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code. |
| MCFLARE | Abstract | C00093 I7090 00 | Monte Carlo Code to Simulate Solar Flare Events and Estimate Probable Doses Encountered on Interplanetary Missions. |
| 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. |
MCNP6.3-EXE
810 | Abstract | C00870 MNYCP 01 | Monte Carlo N-Particle Transport Code System. |
MCNP-DSP-EXE
810 | Abstract | C00699 MNYCP 01 | Monte Carlo N-Particle Transport Code System with Digital Signal Processing based on MCNP4A. |
MCNPX-POLIMI-EXE
810 | Abstract | C00791 MNYCP 01 | Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities. |
| MCRAC | Abstract | C00562 IBMPC 00 | Multiple Cycle Reactor Analysis Code. |
| MCRTOF | Abstract | C00435 FM200 00 | Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region. |
| MCRTOF | Abstract | C00435 I0360 00 | Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region. |
MCUNED
810 | Abstract | C00804 PCX86 00 | MCNPX Extension for Using Light Ion Evaluated Nuclear Data Library. |
| MCVIEW | Abstract | P00202 FM780 00 | View Factor Calculation for Three-Dimensional Geometries. |
| MECC-7 | Abstract | C00156 I0360 00 | Medium-Energy Intranuclear Cascade Code System. |
| MEDUSA-PIJ | Abstract | C00349 F2307 00 | One-Dimensional Lagrangian Code for Plasma Hydrodynamic Analysis of a Fusion Pellet Driven by Ion Beams. |
| MEGA | Abstract | C00839 MNYCP 00 | MEGA: Mechanistic and Engineering Fission Gas Release Prediction Model for UO2 Fuel |
| 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. |
| MERCURE 4-82 | Abstract | C00142 I3033 00 | Three-Dimensional Code System for Integrating Multigroup Line-of-Sight Attenuation Kernels by Monte Carlo Techniques. |
| MESA | Abstract | P00223 I3033 00 | Non-Linear Least Squares Spectral Analysis. |
| MESODIF-II | Abstract | C00498 D0780 00 | A Variable Trajectory Plume Segment Model to Assess Ground-Level Air Concentrations and Depositions of Routine Effluent Releases from Nuclear Power Facilities. |
| MESOI | Abstract | C00497 D0780 00 | Interactive Mesoscale Lagrangian Puff Dispersion Model with Deposition and Decay. |
| MESORAD 1.4 | Abstract | C00677 D0VAX 00 | Code System for Emergency Response Dose Assessment. |
| MESYST | Abstract | C00706 MNYWS 00 | Code System to Simulate 3D Tracer Dispersion in Atmosphere. |
| METD | Abstract | P00197 DGMV1 00 | Computer Code Systems for Use with Meteorological Data. |
| METD | Abstract | P00197 I3033 00 | Computer Code Systems for Use with Meteorological Data. |
| MEVDP | Abstract | C00157 C6600 00 | Primary Radiation Transport Code System - Complex Geometry - Computerized Anatomical Model Man. |
| MGA8 | Abstract | P00542 MNYCP 00 | Code System to Determine Pu Isotope Abundances from Multichannel Analyzer Gamma Spectra. |
| MGCLIB | Abstract | D00118 FM380 00 | 137 and 26 Neutron Multigroup Cross Section Library with the Bondarenko Type Shielding Table. |
| MICAP | Abstract | P00261 I3033 00 | A Monte Carlo Code System for Analysis of Ionization Chamber Responses. |
| 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. |
| MIGROS3 | Abstract | P00265 I0370 00 | A Code for the Generation of Group Constants for Reactor Calculations from Neutron Nuclear Data in KEDAK Format. |
| MILDOS | Abstract | C00398 C0000 00 | Calculation of Radiation Doses from Uranium Recovery Operations. |
| MILDOS-AREA | Abstract | C00608 IBMPC 00 | Calculation of Radiation Dose from Uranium Recovery Operations for Large-Area Sources. |
| MINET | Abstract | P00490 CY000 00 | Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis. |
| 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. |
| MINTEQ | Abstract | P00494 DVX11 00 | Code System to Model Aqueous Geochemical Equilibria. |
| 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. |
| MKENO-DAR | Abstract | C00513 FM380 00 | Direct Angular Representation Monte Carlo Code for Criticality Safety Analysis |
| MMCR | Abstract | C00441 FM200 00 | Multigroup Monte Carlo Neutron and Photon Transport Code. |
| MMRW | Abstract | M00018 MNYCP 00 | Canadian and Early British Energy Reports on Nuclear Reactor Theory (1940-1946). |
| MMRW-BOOKS | Abstract | M00020 MNYCP 00 | MMRW-BOOKS: Legacy books on slowing down, thermalization, particle transport theory, random processes in reactors. |
| MMS3D | Abstract | C00841 MNYCP 00 | Method of Manufactured Solutions for 3D one-group SN Equations with escalating order of non-smoothness |
| MOCA | Abstract | C00590 IPCAT 00 | Monte Carlo Criticality Code System for Hexagonal Geometries. |
| MOCUP | Abstract | P00365 DALPU 00 | MCNP/ORIGEN Coupling Utility Programs. |
| MODEL | Abstract | C00329 I3033 00 | Models of Trapped Proton and Electron Environments for Solar Maximum and Minimum. |
| MOMENT I | Abstract | C00188 U1108 00 | Moments Method Neutron Transport Code System. |
| MOMGEM-MOMDIS | Abstract | C00085 I7090 00 | Moments Method Reconstruction of Scattered Gamma-Ray Distributions. |
MONK 6.3
FEDC | Abstract | C00393 I3033 00 | A General Purpose Monte Carlo Neutronics Code System. |
| MONTEBURNS 2.0 | Abstract | P00455 MNYCP 02 | Automated, Multi-Step Monte Carlo Burnup Code System. |
| 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. |
| MORECA | Abstract | P00411 PC386 00 | Computer Code System for Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup. |
| MORN | Abstract | P00062 I0360 00 | Calculation of the Response of Sodium Iodide Crystals to Gamma Rays. |
| MORSE-ALB | Abstract | C00394 FM200 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-ANSI STD. | Abstract | C00127 I3675 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-B | Abstract | C00368 I0370 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-C | Abstract | C00431 C7600 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CG | Abstract | C00203 C0000 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CG | Abstract | C00203 CY000 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CG | Abstract | C00203 D0VAX 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CG | Abstract | C00203 I0360 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CG | Abstract | C00203 U0000 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CGA | Abstract | C00474 ALLCP 03 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSEC-SP2 | Abstract | P00142 H6000 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CV | Abstract | C00535 HM280 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-E | Abstract | C00258 I0360 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-EMP | Abstract | C00588 IBMPC 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-H | Abstract | C00471 I3081 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-L | Abstract | C00261 C6600 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-SGC | Abstract | C00277 C7600 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MOSRA-LIGHT | Abstract | P00505 MNYWS 00 | High-Speed Three-Dimensional Nodal Diffusion Code System. |
| MOXY-MOD32 | Abstract | P00385 I0360 00 | BWR Core Heat Transfer Code System. |
MRIPP 1.0
810 | Abstract | C00655 PC386 00 | Magnetic Resonance Image Phantom Code System to Calibrate in vivo Measurement Systems. |
| MRSPAK | Abstract | P00212 DVX11 00 | A Code System To Generate a Text File Containing Combinatorial Geometry Data Corresponding to PADL2 Geometry. |
| MSM-SOURCE | Abstract | P00369 MNYCP 00 | Code System for Generation of Input Data for MCNP. |
| MTR_PC 2.6 | Abstract | C00674 PC386 00 | Modular Code System for Neutronics, Thermalhydraulics and Shielding Calculations. |
| MULTI-KENO2 | Abstract | C00492 FM380 00 | A Monte Carlo Code System for Criticality Safety Analysis. |
| MUP2 | Abstract | P00289 I3090 00 | A Program to Calculate Fast Neutron Data for Medium-Heavy Nuclei. |
| MURE V2-SMURE | Abstract | C00764 MNYWS 01 | Serpent - MCNP Utility for Reactor Evolution. |
| MURLI | Abstract | C00378 DP011 00 | Integral Transport Theory Code System for Thermal Reactor Lattice Cell Calculation. |
| MUSCAT | Abstract | C00281 I0360 00 | Calculation of Neutron Currents in Spherical and Cylindrical Cavities by Means of View Factors. |
| MUSPALB | Abstract | C00171 ICL00 00 | Albedo Calculation of Multigroup Spectra of Neutrons Transmitted Through Multilayer Slab Shielding. |
| MUXS | Abstract | P00187 I3033 00 | Generator of Multigroup Cross Sections for Charged Particle Transport Problems. |
| MVP-GMVP II | Abstract | C00739 MNYCP 00 | General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods. |
| MYRA | Abstract | C00056 C0000 00 | Calculation of Shipping Costs and Cask Designs for Irradiated Fuel Elements. |
| MYRA | Abstract | C00056 I7090 00 | Calculation of Shipping Costs and Cask Designs for Irradiated Fuel Elements. |