Online Catalog
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Note: RESTRICTIONS APPLY TO SOME PACKAGES -
810 -- US DOE 10CFR810 Jurisdiction
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Packages starting with A
Package NameAbstractRSICC TapelistTitle
ABAREXAbstractP00248 MNYCP 01Neutron Spherical Optical-Statistical Model Code System.
ABBN-90AbstractD00182 MNYCP 00Multigroup Constant Set for Calculation of Neutron and Photon Radiation Fields and Functionals, Including the CONSYST2 Program.
ABLEIT-TRANSAbstractP00247 C0175 00Error Propagation Analysis for Burnup Calculation.
ACAB-2008AbstractC00758 MNYCP 01Activation Abacus Inventory Code System for Nuclear Applications.
ACATAbstractP00257 FM380 00Monte Carlo Simulation of Atomic Collisions in Amorphous Targets in the Binary Collision Approximation.
ACDOS3AbstractC00442 C7600 00Calculation of Activities and Dose Rates Produced by Neutron Activation.
ACFAAbstractC00478 I3033 00A Versatile Activation Code for Coolant and Structural Materials.
ACHILLESAbstractM00019 MNYCP 00Heat Transfer in PWR Core During LOCA Reflood Phase.
ACOHAbstractC00191 I3675 00Aerojet COHORT Monte Carlo Code System.
ACORNSAbstractP00264 IBMPC 01Analysis of Correlations Used in Neutron Spectrometry.
ACRA-IIAbstractC00213 I0360 00Kernel Integration Code System for Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident.
ACRA-TRITAbstractC00283 I0360 00The Tritium Version of ACRA-II, Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident.
ACROAbstractC00354 I0360 00Calculation of Organ Dose from Acute or Chronic Inhalation and Ingestion of Radionuclides.
ACT-ARAAbstractC00372 CYXMP 00Code System for the Calculation of Changes in Radiological Source Terms with Time.
ACTIVAbstractP00590 I0370 00Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment.
ACTIV87AbstractD00169 ALLCP 00Fast Neutron Activation Cross Section File.
ACTIV-PCAbstractP00287 IBMPC 00A Program to Process Gamma or X-ray Spectra.
ACTL82AbstractD00069 ALLCP 01Evaluated Neutron Activation Cross-Section Library.
ACTV-F/HAbstractD00155 ALLCP 00Neutron Activation Cross Section Library for Fusion Reactor Design.
ACTV-FUS/INTAbstractD00170 ALLCP 00International Library of Neutron Activation Cross-Section Data for Fusion Reactor Application.
ADASAGEAbstractP00426 IBMPC 00Ada Application Development System, Versions 4.02, 4.0 and 3.1.
ADEFTA 4.1AbstractP00543 MNYCP 01Atomic Densities for Transport Analysis Script.
ADENAAbstractP00190 C0000 00Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra.
ADENAAbstractP00190 I3033 00Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra.
ADJMOMAbstractC00212 I3675 00Adjoint Moments Method Gamma-Ray Transport Code System.
ADLER IIIAbstractP00058 I0360 00A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters.
ADOAbstractC00189 I3675 00Aerojet Discrete Ordinates Calculational System.
ADS-LIB/V2.0AbstractD00250 MNYCP 00Test Library for Accelerator Driven Systems V2.0
AGDATAAbstractD00127 I0360 00Two Agricultural Production Data Libraries (AGDATC and AGDATG) for Dose and Risk Assessment Models.
AIR DATAAbstractD00014 I0360 00Sample Input to ANISN for Calculation of Neutron and Secondary Gamma-Ray Transport in Air.
AIRBORNEAbstractC00263 I0360 00Airborne Contaminants Dispersion Code.
AIRDIFAbstractC00360 C6600 00A Two-Dimensional Atmospheric Radiation Diffusion Code.
AIRDOS-PCAbstractC00551 IBMPC 00Clean Air Act Compliance Software for Personal Computers. See C00542/CAP-88.
AIREKMOD-RRAbstractP00588 D0VAX 00A Point Kinetic Computer Code for Reactivity Transient Analysis in Nuclear Research Reactors AIREKMOD-RR User's Guide.
AIREMAbstractC00242 I3691 00Calculation of Doses, Population Doses, and Ground Depositions Due to Atmospheric Emissions of Radionuclides.
AIRFEWGAbstractD00049 I0360 00Results 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.
AIRGAMMAAbstractC00567 FM380 00A Program For The Calculation Of External Exposure To Gamma Rays From A Radioactive Cloud.
AIRSCATAbstractC00341 DP010 00Calculation of Dose Rate for Gamma-Rays Scattered in Air.
AIRTRANSAbstractC00110 I3675 00Monte Carlo Time and Energy-Dependent Three-Dimensional Radiation Transport Code.
AISITE IIAbstractC00286 I0360 00Reactor Siting Code System.
AKERNAbstractC00190 C0000 00Aerojet Point Kernel Integration Calculational System.
AKERNAbstractC00190 U1108 00Aerojet Point Kernel Integration Calculational System.
AKTIVAbstractC00339 I0360 00An Evaluation of Activity, Afterheat and Biological Hazard Potential of Stainless Steel Structures in Fusion Reactor Blankets.
ALARA 2.7.8AbstractC00723 MNYCP 00Code System for Analytic and Laplacian Adaptive Radioactivity Analysis.
ALARM-B2AbstractP00218 I0360 00A Computer Code System for Analysis of a Large Break LOCA of a BWR.
ALBEDO/ALBEZAbstractC00555 IBMPC 00Calculates Attenuation of Radiation in Single and Double Bends.
ALBEDO-DATAAbstractD00224 MNYCP 00KSU Neutron Albedo Data.
ALBEMOAbstractC00268 C6600 00Albedo Monte Carlo Code System.
ALDOSEAbstractC00577 IBMPC 00Dose Calculation for Alpha Disc Source.
ALEPH-LIB-JEFF3.1AbstractD00230 MNYCP 00ACE Format Neutron Cross Section Library based on JEFF3.1.
ALGAM-97AbstractC00152 I3675 00Monte Carlo Estimation of Internal Dose from Gamma-Ray Sources in a Phantom Man.
ALICE2012AbstractP00550 PCX86 04Statistical Model Code System to Calculate Particle Spectra from HMS Precompound Nucleus Decay.
ALKASYS-PCAbstractC00558 IBMPC 00A Computer Program For Studies of Rankine-Cycle Space Nuclear Power Systems.
ALPHA-MAbstractP00169 I0360 00Least-Squares Resolution of Gamma-Ray Spectra in Environmental Samples.
ALPHNAbstractC00612 IBMPC 00Code System for Calculating (alpha,n) Neutron Production in Canisters of High-Level Waste.
ALPSAbstractP00144 F2307 00Alpha Spectrum Analysis Code System.
AMARAAbstractP00079 I3675 00Nuclear Data Adjustment Using Lagrange's Multipliers Method.
AMCAbstractC00090 I3675 00Monte Carlo Albedo Code for Neutron and Capture Gamma-Ray Distributions in Rectangular Concrete Ducts.
AMPAbstractC00793 PCX86 00Advanced Multi-Physics.
AMPX01AbstractD00027 I3675 02126-Group Coupled Neutron and Gamma-Ray Transport Cross-Section Data Generated by AMPX.
AMPX-77AbstractP00315 ALLMF 01Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B.
AMUSEAbstractP00028 C6600 00Gamma-Ray Spectra Unfolding Code.
ANAAbstractP00356 IBMPC 00Code System for Gamma-Ray Spectra Analyses.
ANGELO-LAMBDAAbstractP00544 MNYCP 01Covariance Matrix Interpolation and Mathematical Verification.
ANIPLO D50AbstractP00213 I0360 00A Digital Computer Program for Plotting Results from Calculations with the Sn Computer Program ANISN.
ANISN-ORNLAbstractC00254 MNYCP 02One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering.
ANISN-PCAbstractC00514 IBMPC 00Multigroup One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering. RSIC recommends CCC-650/DOORS3.2a for most applications.
ANITA-2000AbstractC00693 MNYCP 00Code System to Calculate Isotope Inventories from Neutron Irradiation for Fusion Applications.
ANITA-4AbstractC00606 MNYCP 01Analysis of Neutron Induced Transmutation and Activation. See ANITA-2000 (CCC-693).
ANL-BPBAbstractM00004 MNYCP 00Argonne National Laboratory Code Center: Benchmark Problem Book.
ANS643AbstractD00129 IBMPC 02ANS-6.4.3 Geometric Progression Gamma-Ray Buildup Factor Coefficients.
ANSIFTAbstractP00077 C6600 00ANSI Standard Fortran Sifting Program.
ANSIFTAbstractP00077 I0360 00ANSI Standard Fortran Sifting Program.
ANSL-VAbstractD00154 ALLCP 01ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies.
ANTE 2AbstractC00131 I3675 00Adjoint Monte Carlo Time-Dependent Neutron Transport Code in Combinatorial Geometry.
APARNA-IIAbstractC00296 I0360 00Integral Transport Theory Code System Based on Discrete Ordinate Representation in Space and Direction-Slab Geometry.
APPLE-2AbstractP00111 FM200 00Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates.
APPLE-2AbstractP00111 I3081 00Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates.
APSAIAbstractP00065 I3691 00Activity Calculations and Plotting of Neutron or Gamma-Ray Spectra Generated by Discrete Ordinates Code System ANISN.
APUD 3.0AbstractC00637 IBMPC 00Code System for Analyzing, Predicting Consequences of, and Guiding the Response to Nuclear Emergencies.
ARCAbstractC00224 C6600 00Aircraft Radiation Transport Code System, Crew Dose Calculation.
ARCON96AbstractC00664 IBMPC 00Code System to Calculate Atmospheric Relative Concentrations in Building Wakes.
AREACAbstractC00438 I3033 00Radiological Emission Analysis Code System.
AREADAbstractP00088 I0360 00Input Data Processor for Transport Codes.
ARMYL-GAbstractC00297 U1106 00Calculation of Transmission Factors for Gamma Rays from Nuclear Explosions.
ARMYL-NAbstractC00298 U1106 00Calculation of Transmission Factors for Neutrons from Nuclear Explosions.
ARRRGAbstractC00404 U1100 00Calculation of Radiation Dose to Man from Radionuclides in the Environment.
ASFIT-VARIAbstractC00336 H0000 00Gamma-Ray Transport Code System for One-Dimensional Finite Systems.
ASFIT-VARIAbstractC00336 IBMPC 00Gamma-Ray Transport Code System for One-Dimensional Finite Systems.
ASOPAbstractC00126 IRISC 00Multigroup One-Dimensional Discrete Ordinates Transport Code System for Shield Optimization.
ASTROSAbstractC00073 I7090 00Calculation of Primary and Secondary Proton Dose Rates in Spheres and Slabs of Tissue.
AT123DAbstractC00417 I0360 00Analytical Transient One-, Two-, and Three-Dimensional Simulation of Waste Transport in an Aquifer System.
ATHENA_2DAbstractP00431 MNYCP 00Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum.
ATM-TOXAbstractC00472 I3033 00An Atmospheric Transport Model for Toxic Substances.
ATTOW-KBAbstractC00132 I0370 00Multigroup Two-Dimensional Removal-Diffusion (Spinney Method) Shielding Code System.
AUS98AbstractC00519 MNYWS 01Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems.
AUTOJOM-JOMREADAbstractP00008 C6600 00Computer Programs to Generate or Check Coefficients for Quadratic Equations Describing 3D Geometries.
AXMIXAbstractP00075 CYXMP 00ANISN Cross Section Code System.
AXMIXAbstractP00075 IRISC 01ANISN Cross Section Code System.
AXMIX-PCAbstractP00297 IBMPC 00ANISN Cross Section Mixing Code System.
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.