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 T
Package NameAbstractRSICC TapelistTitle
TACT-IIIAbstractC00447 I3033 00Calculation of the Transport of Radioactivity from a Reactor Core.
TALYS-1.2AbstractP00548 PC586 01Nuclear Model Code System for Analysis and Prediction of Nuclear Reactions and Generation of Nuclear Data.
TAM3AbstractP00308 IBMPC 00Demonstrates Monte Carlo Sensitivity and Uncertainty Analysis.
TART2012AbstractC00638 MNYCP 07Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code System.
TASKAbstractC00184 I0360 00Generalized One-Dimensional Radiation Transport and Diffusion Kinetics Code System.
TDAAbstractC00180 MNYWS 01A Time-Dependent, Multigroup, One-Dimensional, Discrete Ordinates Transport Code System.
TDFAbstractD00162 ALLCP 00Thermonuclear Data File.
TDOWN-IVAbstractP00172 H6000 00A Code System to Generate Composition- and Spatially-Dependent Neutron Cross Sections for Multigroup Neutronics Analysis.
TDTAbstractC00256 I0360 00Generalized One-Dimensional Multigroup Time-Dependent Transport and Diffusion Kinetic Code System.
TDTORTAbstractC00709 MNYWS 00Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System.
TECALCAbstractP00074 DP010 00Interactive Calculation of Compton Coherent and Photoelectric Mass Attenuation Coefficients for Photons (E<1 MeV), and the Mass Absorption Coefficient for Known Materials.
TEMACAbstractP00468 D0VAX 00Top Event Matrix Analysis Code System.
TEMPEST-2AbstractP00558 I0360 00Thermalization Program for Neutron Spectra and MultiGroup Cross-Sections.
TEMPEST-BNWAbstractP00559 C7600 00Transient 3-D Thermohydraulics for FBR.
TENDL-2008-ACEAbstractD00243 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2010-ACEAbstractD00248 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2011-ACEAbstractD00252 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2012-ACEAbstractD00266 MNYCP 00TALYS-Based Cross Section Library for Use with MCNPX.
TERFOC-NAbstractC00596 MFMWS 00Terrestrial Food-Chain Model for Normal Operations.
TESSAbstractC00215 C3600 00Multigroup Discrete Ordinates Code System for Slab and Spherical Geometries.
THACT-RRAbstractP00587 D0VAX 00Analysis of Thermal Hydraulics Transients in Research Reactor Core.
THERMGAMAbstractD00140 ALLCP 00Prompt Gamma Rays from Thermal-Neutron Capture.
THERMOS-OTAAbstractP00107 C0173 00Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders.
THERMOS-OTAAbstractP00107 C0740 00Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders.
THERMOS-OTAAbstractP00107 U1108 00Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders.
THIDA-2AbstractC00410 FM380 00Code System for the Calculation of Transmutation, Activation, Decay Heat and Dose Rate in Fusion Reactors.
THRUSHAbstractP00276 CYXMP 00Calculates Thermal Neutron Scattering Kernel.
THTAbstractC00480 I0360 00Three-Dimensional Neutron Coarse Mesh Code System to Evaluate Average Bundle Fluxes and Power in Light Water Reactors.
THYDE-B1/MOD2AbstractP00553 FM200 00Computer Code for the Analysis of Small-Break Loss-of-Coolant Accident of Boiling Water Reactors.
THYDE-P2AbstractP00554 FV100 00Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
TIBSOAbstractC00512 MNYCP 00Code System to Calculate Production and Migration of Radionuclides in Nuclear Reactor Systems.
TIMEDAbstractC00292 I0360 00Calculation of Cumulated Activity of a Radionuclide in the Organs of the Human Body at a Given Time After Deposition.
TIMEXAbstractC00274 C7600 00One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering.
TIMEXAbstractC00274 CY000 00One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering.
TIMEXAbstractC00274 U1106 00One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering.
TIMOC-72AbstractC00144 I0370 00Monte Carlo Three-Dimensional Neutron Transport Code System.
TIMOC-ESPAbstractC00432 U1110 00System for Generating and Analyzing Time Dependent Radiation Transport Results by Monte Carlo.
TIMS-1AbstractP00163 D0780 00Processing Code System for Production of Group Constants of Heavy Resonant Nuclei.
TIMS-1AbstractP00163 FM200 00Processing Code System for Production of Group Constants of Heavy Resonant Nuclei.
TIRION 4AbstractC00395 I3033 00A Program for Calculating Consequences of a Release of Radioactive Material to the Atmosphere.
TITAN 1.29AbstractC00759 PCX86 04A Three-Dimensional Deterministic Radiation Transport Code System.
TMMSAbstractC00246 I0360 00Gamma-Ray Penetration Shielding Code System, Transmission Matrix Method.
TNG1AbstractP00298 D6220 00A Multistep Statistical Model Based on the Hauser-Feshbach Theory For The Evaluation Of Neutron Data.
TORACAbstractP00459 C0170 00Code System to Calculate Tornado-Induced Flow Material Transport.
TOTEM-3AbstractP00603 I0370 00Demand Assessment for Nuclear Power Plants and Conventional Power Plants.
TOXRISKAbstractC00692 CDCMF 00Code System for Toxic Gas Accident Analysis.
TP1AbstractC00465 I3033 00A Computer Code System for the Calculation of Reactivity and Kinetic Parameters by One-Dimensional Neutron Transport Perturbation Theory.
TP2AbstractC00470 I3033 00A Computer Program for the Calculation of Reactivity and Kinetic Parameters by Two-Dimensional Neutron Transport Perturbation Theory.
TPASGAM 85AbstractD00088 ALLCP 04Radioactive Decay Library of Gamma-Ray Energies, Branching Ratios, and Cross Sections.
TPASSAbstractP00164 DP010 00A Gamma-Ray Spectral Data-Reduction and Analysis Code System.
TPHEXAbstractC00421 C0173 00Transmission Probability Code System for Calculating Neutron Flux Distributions in Hexagonal Geometry.
TPHEXAbstractC00421 CYXMP 00Transmission Probability Code System for Calculating Neutron Flux Distributions in Hexagonal Geometry.
TPTRIAAbstractC00550 I3083 00A Computer Program for the Reactivity and Kinetic Parameters for Two-Dimensional Triangular Geometry by Transport Perturbation Theory.
TRAC-BD1
USSO
AbstractP00488 C0176 00Code System for Best-Estimate Analysis of LOCA in BWR.
TRAC-PF1
USSO
AbstractP00481 IBMPC 00Best-Estimate Analysis PWR LOCA.
TRAC-PF1/EN MOD3AbstractP00477 PCX86 01Code System for Coupled 3D Neutronics-Thermalhydraulics Calculations.
TRANSHEXAbstractC00449 U1108 00Two-dimensional Multigroup Collision Probability Code System for Hexagonal Geometry.
TRANSMITAbstractD00020 I0360 00Experimental Neutron Transmission Data Used to Test Total Cross Sections.
TRANSPORTAbstractC00244 C6600 00Charged Particle Beam Transport Systems Design Code System (First- and Second-Order Matrix Multiplication).
TRANSPORTAbstractC00244 I0360 00Charged Particle Beam Transport Systems Design Code System (First- and Second-Order Matrix Multiplication).
TRANSX 2.15AbstractP00317 MFMWS 01Code system to produce neutron, photon, and particle transport tables for discrete-ordinates and diffusion codes from cross sections in MATXS format.
TRANSX-CTRAbstractP00206 CY000 00Interfaces MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis.
TRANZITAbstractC00172 C7600 00Multigroup Time-Dependent Discrete Ordinates Radiation Transport Code System in (rho,z) Cylindrical Geometry.
TRAPPAbstractC00205 I3691 00Transport of Alpha Particles and Protons with all Nuclear Reaction Products Neglected.
TRAXAbstractP00280 C0720 00A Program For Optics of Curved Crystal Neutron Spectrometers.
TRD-3AbstractC00362 I3033 00Two-Dimensional Removal-Diffusion Neutron Shielding Code System.
TRECOAbstractC00116 I3675 00An Orbital Integration Estimation of Trapped Radiation.
TR-EDBAbstractD00198 IBMPC 00Test Reactor Embrittlement Data Base, Version 1.
TREEDEAbstractC00326 C0000 00Monte Carlo Neutron Transport Code System Based on the Track Rotation Estimator.
TRG-SGDAbstractC00025 C0000 00Calculation of Secondary Gamma-Ray Dose Rate from a Nuclear Weapon Detonation-Monte Carlo Method.
TRIDENTAbstractC00293 C7600 00Two-Dimensional Multigroup Discrete Ordinates Transport Code System-(x,y) and (r,z) Geometries.
TRIDENTAbstractC00293 I0360 00Two-Dimensional Multigroup Discrete Ordinates Transport Code System-(x,y) and (r,z) Geometries.
TRIDENT-CTRAbstractC00377 C0000 00Two-Dimensional x-y and r-z Geometry Multigroup Transport Code System for Large Toroidal Reactors.
TRIGAPAbstractC00600 IBMPC 00A Computer Code for TRIGA Type Reactors.
TRIGLAVAbstractP00495 PC586 00Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor.
TRIGONAbstractC00290 U1108 00Two-Dimensional Multigroup Diffusion Code System-Trigonal or Hexagonal Mesh.
TRIPLETAbstractC00230 C6600 00Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System.
TRIPLETAbstractC00230 C7600 00Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System.
TRIPLETAbstractC00230 I0360 00Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System.
TRIPOLI-4 8.1
OECD
AbstractC00806 MNYCP 003D General Purpose Continuous Energy Monte Carlo Transport Code.
TRIPOLI-4 9S
OECD
AbstractC00815 MNYCP 00Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo Transport Calculation.
TRIPOSAbstractC00537 CY00I 00Monte Carlo Ion Transport Analysis Code.
TRISTANAbstractC00511 HM280 00Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System.
TRISTAN-IJSAbstractP00537 IBMPC 00Steady-State Axial Temperature and Flow Velocity in Triga Channel.
TRITACAbstractC00560 D8810 00A Three-Dimensional Transport Code For Eigenvalue Problems Using The Diffusion Synthetic Acceleration Method.
TRUMPAbstractP00522 MNYCP 01Code System for Transient and Steady-State Temperature Distribution in Multidimensional Systems.
TSORTAbstractP00486 IBMPC 00Automated Technique for Nuclear Plant Training Task Assignment.
TURBINAAbstractP00604 I0370 00Reheat Steam Turbine Generator Design with Preheater and Condenser.
TWOTRANAbstractC00195 C6600 00Two-Dimensional Discrete Ordinates. We recommend CCC-547/TWODANT-SYS.
TWOTRAN IIAbstractC00222 C7600 00Two-Dimensional Multigroup Discrete Ordinates Transport C System in (x,y), (r,theta), and (r,z) Geometries.
TWOTRAN IIAbstractC00222 I3691 00Two-Dimensional Multigroup Discrete Ordinates Transport C System in (x,y), (r,theta), and (r,z) Geometries.
TWOTRAN-SPHEREAbstractC00129 C6600 00Multigroup Two-Dimensional Discrete Ordinates Transport Code System in Spherical Geometry.
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.