| Packages starting with A |
| Package Name | Abstract | RSICC Tapelist | Title |
| AARE-V1.0 | Abstract | C00846 MNYCP 00 | Activation in Accelerator Radiation Environments |
| ABAREX | Abstract | P00248 MNYCP 01 | Neutron Spherical Optical-Statistical Model Code System. |
| ABBN-90 | Abstract | D00182 MNYCP 00 | Multigroup Constant Set for Calculation of Neutron and Photon Radiation Fields and Functionals, Including the CONSYST2 Program. |
| ABLEIT-TRANS | Abstract | P00247 C0175 00 | Error Propagation Analysis for Burnup Calculation. |
| ACAB-2008 | Abstract | C00758 MNYCP 01 | Activation Abacus Inventory Code System for Nuclear Applications. |
| ACAT | Abstract | P00257 FM380 00 | Monte Carlo Simulation of Atomic Collisions in Amorphous Targets in the Binary Collision Approximation. |
| ACDOS3 | Abstract | C00442 C7600 00 | Calculation of Activities and Dose Rates Produced by Neutron Activation. |
| ACFA | Abstract | C00478 I3033 00 | A Versatile Activation Code for Coolant and Structural Materials. |
| ACOH | Abstract | C00191 I3675 00 | Aerojet COHORT Monte Carlo Code System. |
| ACORNS | Abstract | P00264 IBMPC 01 | Analysis of Correlations Used in Neutron Spectrometry. |
| ACRA-II | Abstract | C00213 I0360 00 | Kernel Integration Code System for Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident. |
| ACRA-TRIT | Abstract | C00283 I0360 00 | The Tritium Version of ACRA-II, Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident. |
| ACT-ARA | Abstract | C00372 CYXMP 00 | Code System for the Calculation of Changes in Radiological Source Terms with Time. |
| ACTIV | Abstract | P00590 I0370 00 | Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment. |
| ACTIV87 | Abstract | D00169 ALLCP 00 | Fast Neutron Activation Cross Section File. |
| ACTIV-PC | Abstract | P00287 IBMPC 00 | A Program to Process Gamma or X-ray Spectra. |
| 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. |
| ADASAGE | Abstract | P00426 IBMPC 00 | Ada Application Development System. |
| ADEFTA 4.1 | Abstract | P00543 MNYCP 01 | Atomic Densities for Transport Analysis Script. |
| 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. |
| ADJMOM | Abstract | C00212 I3675 00 | Adjoint Moments Method Gamma-Ray Transport Code System. |
| ADLER III | Abstract | P00058 I0360 00 | A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters. |
| ADO | Abstract | C00189 I3675 00 | Aerojet Discrete Ordinates Calculational System. |
| ADS-LIB/V2.0 | Abstract | D00250 MNYCP 00 | Test Library for Accelerator Driven Systems V2.0 |
ADVANTG 3.2.0
810 | Abstract | C00854 PCX86 00 | AutomateD VAriaNce reducTion Generator |
| ADVANTG 3.2.1 | Abstract | C00854 PCX86 01 | AutomateD VAriaNce reducTion Generator |
| 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. |
| AIRBORNE | Abstract | C00263 I0360 00 | Airborne Contaminants Dispersion Code. |
| AIRDIF | Abstract | C00360 C6600 00 | A Two-Dimensional Atmospheric Radiation Diffusion Code. |
| AIRDOS-PC | Abstract | C00551 IBMPC 00 | Clean Air Act Compliance Software for Personal Computers. |
| AIREKMOD-RR | Abstract | P00588 D0VAX 00 | Reactivity Transients in Nuclear Research Reactors |
| AIREKMOD-RR | Abstract | P00588 PCX86 01 | Reactivity Transients in Nuclear Research Reactors |
| AIREM | Abstract | C00242 I3691 00 | Calculation of Doses, Population Doses, and Ground Depositions Due to Atmospheric Emissions of Radionuclides. |
| 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. |
| AIRGAMMA | Abstract | C00567 FM380 00 | A Program For The Calculation Of External Exposure To Gamma Rays From A Radioactive Cloud. |
| AIRSCAT | Abstract | C00341 DP010 00 | Calculation of Dose Rate for Gamma-Rays Scattered in Air. |
| AIRTRANS | Abstract | C00110 I3675 00 | Monte Carlo Time and Energy-Dependent Three-Dimensional Radiation Transport Code. |
| AISITE II | Abstract | C00286 I0360 00 | Reactor Siting Code System. |
| AKERN | Abstract | C00190 C0000 00 | Aerojet Point Kernel Integration Calculational System. |
| AKERN | Abstract | C00190 U1108 00 | Aerojet Point Kernel Integration Calculational System. |
| AKTIV | Abstract | C00339 I0360 00 | An Evaluation of Activity, Afterheat and Biological Hazard Potential of Stainless Steel Structures in Fusion Reactor Blankets. |
| ALARA 2.7.8 | Abstract | C00723 MNYCP 00 | Code System for Analytic and Laplacian Adaptive Radioactivity Analysis. |
| ALARM-B2 | Abstract | P00218 I0360 00 | A Computer Code System for Analysis of a Large Break LOCA of a BWR. |
| ALBEDO/ALBEZ | Abstract | C00555 IBMPC 00 | Calculates Attenuation of Radiation in Single and Double Bends. |
| ALBEDO-DATA | Abstract | D00224 MNYCP 00 | KSU Neutron Albedo Data. |
| ALBEMO | Abstract | C00268 C6600 00 | Albedo Monte Carlo Code System. |
| ALDOSE | Abstract | C00577 IBMPC 00 | Dose Calculation for Alpha Disc Source. |
| ALEPH-LIB-JEFF3.1 | Abstract | D00230 MNYCP 00 | ACE Format Neutron Cross Section Library based on JEFF3.1. |
| ALGAM-97 | Abstract | C00152 I3675 00 | Monte Carlo Estimation of Internal Dose from Gamma-Ray Sources in a Phantom Man. |
| ALICE2017 | Abstract | P00550 PCX86 06 | Statistical Model Code System to Calculate Particle Spectra from HMS Precompound Nucleus Decay. |
| ALKASYS-PC | Abstract | C00558 IBMPC 00 | A Computer Program For Studies of Rankine-Cycle Space Nuclear Power Systems. |
| ALPHA-M | Abstract | P00169 I0360 00 | Least-Squares Resolution of Gamma-Ray Spectra in Environmental Samples. |
| ALPHN | Abstract | C00612 IBMPC 00 | Code System for Calculating (alpha,n) Neutron Production in Canisters of High-Level Waste. |
| AMARA | Abstract | P00079 I3675 00 | Nuclear Data Adjustment Using Lagrange's Multipliers Method. |
| AMC | Abstract | C00090 I3675 00 | Monte Carlo Albedo Code for Neutron and Capture Gamma-Ray Distributions in Rectangular Concrete Ducts. |
| AMP | Abstract | C00793 PCX86 00 | Advanced Multi-Physics. |
| AMPX01 | Abstract | D00027 I3675 02 | Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
| AMPX-77 | Abstract | P00315 ALLMF 01 | Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
| AMUSE | Abstract | P00028 C6600 00 | Gamma-Ray Spectra Unfolding Code. |
| ANA | Abstract | P00356 IBMPC 00 | Code System for Gamma-Ray Spectra Analyses. |
| ANGELO-LAMBDA | Abstract | P00544 MNYCP 01 | Covariance Matrix Interpolation and Mathematical Verification. |
| ANIPLO D50 | Abstract | P00213 I0360 00 | A Digital Computer Program for Plotting Results from Calculations with the Sn Computer Program ANISN. |
| ANISN-ORNL | Abstract | C00254 MNYCP 02 | One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering. |
| ANISN-PC | Abstract | C00514 IBMPC 00 | Multigroup One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering. |
| ANITA-2000 | Abstract | C00693 MNYCP 00 | Analysis of Neutron Induced Transmutation and Activation. |
| ANITA-4 | Abstract | C00606 MNYCP 01 | Analysis of Neutron Induced Transmutation and Activation. |
| ANL-BPB | Abstract | M00004 MNYCP 00 | Argonne National Laboratory Code Center: Benchmark Problem Book. |
| ANS643 | Abstract | D00129 IBMPC 02 | Geometric Progression Gamma-Ray Buildup Factor Coefficients. |
| ANSIFT | Abstract | P00077 C6600 00 | ANSI Standard Fortran Sifting Program. |
| ANSIFT | Abstract | P00077 I0360 00 | ANSI Standard Fortran Sifting Program. |
| ANSL-V | Abstract | D00154 ALLCP 01 | ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies. |
| ANTE 2 | Abstract | C00131 I3675 00 | Adjoint Monte Carlo Time-Dependent Neutron Transport Code in Combinatorial Geometry. |
| APARNA-II | Abstract | C00296 I0360 00 | Integral Transport Theory Code System Based on Discrete Ordinate Representation in Space and Direction-Slab Geometry. |
| APPLE-2 | Abstract | P00111 FM200 00 | Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates. |
| APPLE-2 | Abstract | P00111 I3081 00 | Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates. |
| APSAI | Abstract | P00065 I3691 00 | Activity Calculations and Plotting of Neutron or Gamma-Ray Spectra Generated by Discrete Ordinates Code System ANISN. |
| APUD 3.0 | Abstract | C00637 IBMPC 00 | Code System for Analyzing, Predicting Consequences of, and Guiding the Response to Nuclear Emergencies. |
| ARC | Abstract | C00224 C6600 00 | Aircraft Radiation Transport Code System, Crew Dose Calculation. |
ARC 11.2892
FEDC | Abstract | C00824 MNYCP 02 | Code System for Analysis of Nuclear Reactors. |
| ARCON96 | Abstract | C00664 IBMPC 00 | Code System to Calculate Atmospheric Relative Concentrations in Building Wakes. |
| AREAC | Abstract | C00438 I3033 00 | Radiological Emission Analysis Code System. |
| AREAD | Abstract | P00088 I0360 00 | Input Data Processor for Transport Codes. |
| ARMYL-G | Abstract | C00297 U1106 00 | Calculation of Transmission Factors for Gamma Rays from Nuclear Explosions. |
| ARMYL-N | Abstract | C00298 U1106 00 | Calculation of Transmission Factors for Neutrons from Nuclear Explosions. |
| ARRRG | Abstract | C00404 U1100 00 | Calculation of Radiation Dose to Man from Radionuclides in the Environment. |
| ART MOD2 | Abstract | P00611 PCX86 00 | Fission Product Migration in Primary System and Containment |
| ASFIT-VARI | Abstract | C00336 H0000 00 | Gamma-Ray Transport Code System for One-Dimensional Finite Systems. |
| ASFIT-VARI | Abstract | C00336 IBMPC 00 | Gamma-Ray Transport Code System for One-Dimensional Finite Systems. |
| ASOP | Abstract | C00126 IRISC 00 | Multigroup One-Dimensional Discrete Ordinates Transport Code System for Shield Optimization. |
| ASTROS | Abstract | C00073 I7090 00 | Calculation of Primary and Secondary Proton Dose Rates in Spheres and Slabs of Tissue. |
| AT123D | Abstract | C00417 I0360 00 | Analytical Transient One-, Two-, and Three-Dimensional Simulation of Waste Transport in an Aquifer System. |
| ATHENA_2D | Abstract | P00431 MNYCP 00 | Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum. |
| ATM-TOX | Abstract | C00472 I3033 00 | An Atmospheric Transport Model for Toxic Substances. |
| ATTOW-KB | Abstract | C00132 I0370 00 | Multigroup Two-Dimensional Removal-Diffusion (Spinney Method) Shielding Code System. |
| AUS98 | Abstract | C00519 MNYWS 01 | Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems. |
| AUTOJOM-JOMREAD | Abstract | P00008 C6600 00 | Computer Programs to Generate or Check Coefficients for Quadratic Equations Describing 3D Geometries. |
| AXMIX-PC | Abstract | P00297 IBMPC 00 | ANISN Cross Section Code System. |