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Packages with Keyword: NEUTRON
Package NameAbstractRSICC TapelistTitle
AARE-V1.0AbstractC00846 MNYCP 00Activation in Accelerator Radiation Environments
ABAREXAbstractP00248 MNYCP 01Neutron Spherical Optical-Statistical Model Code System.
ACAB-2008AbstractC00758 MNYCP 01Activation Abacus Inventory Code System for Nuclear Applications.
ACDOS3AbstractC00442 C7600 00Calculation of Activities and Dose Rates Produced by Neutron Activation.
ACOHAbstractC00191 I3675 00Aerojet COHORT Monte Carlo Code System.
ADOAbstractC00189 I3675 00Aerojet Discrete Ordinates Calculational System.
ADS-LIB/V2.0AbstractD00250 MNYCP 00Test Library for Accelerator Driven Systems V2.0
AIRDIFAbstractC00360 C6600 00A Two-Dimensional Atmospheric Radiation Diffusion Code.
AIRTRANSAbstractC00110 I3675 00Monte Carlo Time and Energy-Dependent Three-Dimensional Radiation Transport Code.
AKERNAbstractC00190 C0000 00Aerojet Point Kernel Integration Calculational System.
AKERNAbstractC00190 U1108 00Aerojet Point Kernel Integration Calculational System.
ALARA 2.7.8AbstractC00723 MNYCP 00Code System for Analytic and Laplacian Adaptive Radioactivity Analysis.
ALBEDO/ALBEZAbstractC00555 IBMPC 00Calculates Attenuation of Radiation in Single and Double Bends.
ALBEMOAbstractC00268 C6600 00Albedo Monte Carlo Code System.
ALPHNAbstractC00612 IBMPC 00Code System for Calculating (alpha,n) Neutron Production in Canisters of High-Level Waste.
AMCAbstractC00090 I3675 00Monte Carlo Albedo Code for Neutron and Capture Gamma-Ray Distributions in Rectangular Concrete Ducts.
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.
ANITA-2000AbstractC00693 MNYCP 00Analysis of Neutron Induced Transmutation and Activation.
ANITA-4AbstractC00606 MNYCP 01Analysis of Neutron Induced Transmutation and Activation.
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.
ARCAbstractC00224 C6600 00Aircraft Radiation Transport Code System, Crew Dose Calculation.
ARMYL-NAbstractC00298 U1106 00Calculation of Transmission Factors for Neutrons from Nuclear Explosions.
ASOPAbstractC00126 IRISC 00Multigroup One-Dimensional Discrete Ordinates Transport Code System for Shield Optimization.
AUS98AbstractC00519 MNYWS 01Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems.
BASACFAbstractP00285 IBMPC 00Bayesian Approach to Spectrum Adjustment with Covariance Filter.
BERMUDAAbstractC00616 FV260 03Discrete Ordinates Code System for Shielding Analysis for Use with Fusion and Fission Reactors.
BISON 1.5AbstractC00464 HM200 00One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System.
BISON-CAbstractC00659 MNYWS 00One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System.
BMC-MGAbstractC00291 C6600 00Multigroup Monte Carlo Neutron and Gamma-Ray Shielding Code System for Plutonium.
BOLD VENTURE IVAbstractC00459 I3033 00A Reactor Analysis Code System.
BOREHOLE-EB6.8-MGAbstractD00268 MNYCP 00Multi-Group Cross-Section Library for Deterministic and Monte Carlo Codes.
BOT3P-5.3AbstractP00530 MNYCP 02Code System for 2D and 3D Mesh Generation and Graphical Display of Geometry and Results for Radiation Transport Codes.
BOXERAbstractC00766 MNYWS 00Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations
BUCORSTAbstractP00339 PC386 00A Code to Prepare Burnup-Dependent Multigroup Nuclear Reactor Source Terms.
BULK_C-12AbstractC00738 PC586 00Code System to Estimate Neutron and Photon Effective Dose Rates from Medium Energy Protons or Carbon Ions Through Concrete or Concrete/Iron.
BULK-IAbstractP00574 PCX86 00Radiation Shielding Tool for Proton Accelerator Facilities.
CAFDATSAbstractP00549 MNYCP 00Converter of Angular Fluxes of DORT, ANISN and TORT Systems.
CALOR95AbstractC00610 MNYWS 00Monte Carlo Code System for Design and Analysis of Calorimeter Systems, Spallation Neutron Source (SNS) Target Systems, etc.
CARMEN SYSTEMAbstractC00487 U1110 00A Code System for Neutronics PWR Calculation by Diffusion Theory with Space-Dependent Feedback Effects.
CARNACAbstractC00238 I3691 00Calculation of Flux and Neutron Spectra in the Case of Criticality Accident.
CASIMAbstractC00265 I0360 00Monte Carlo Simulation of Transport of Hadron Cascades in Bulk Matter.
CASTHYAbstractP00316 FM000 00Statistical Model Calculation for Neutron Cross Sections and Capture Gamma-Ray Spectra.
CAVEATAbstractC00169 I3675 00General Purpose Monte Carlo Time-Dependent Radiation Transport Code in Complex Geometry.
CCRMNAbstractP00366 MNYCP 00Monte Carlo Simulation of the Coupled Transport of Electrons and Photons.
CDRAbstractC00182 C6600 00A Constant Dose Range Code System, Using the LANL-NWEF Neutron-Gamma-Ray Air Flux Tape.
CDRAbstractC00182 I0360 00A Constant Dose Range Code System, Using the LANL-NWEF Neutron-Gamma-Ray Air Flux Tape.
CINDER 1.05AbstractC00755 PC586 00Code System for Actinide Transmutation Calculations
CLESAbstractD00233 MNYCP 00Cross Section Library of Moderator Materials for Low-Energy Neutron Sources.
CNCSN 2009AbstractC00726 PCX86 01One, Two- and Three-Dimensional Coupled Neutral and Charged Particle SN Parallel Multi-Threaded Code System.
COHORT-IIAbstractC00198 I7094 00General Purpose Monte Carlo Radiation Transport Code System.
COMBINE-PCAbstractP00286 IBMPC 00Code System to Compute Neutron Spectra and ENDF/B Version 5 Based Multigroup Neutron Constants.
COMNUC3BAbstractP00302 CYXMP 00A Compound Nucleus Analysis Program.
COMPARAbstractP00240 C0170 00Compares Multigroup Cross Sections Generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS.
COMPRASHAbstractC00072 I3675 00Spinney Removal-Diffusion Shielding Code.
CRESOAbstractP00184 I3081 00Resonance Data-Handling Code System.
CRYSTAL BALLAbstractC00233 C6600 00Code System for Determining Neutron Spectra from Activation Measurements.
CRYSTAL BALLAbstractC00233 I0360 00Code System for Determining Neutron Spectra from Activation Measurements.
CYGNUS-C SPHEREAbstractC00232 I0360 00Monte Carlo Neutron Transport Code System in Spherical Geometry.
DANTSYS 3.0AbstractC00547 MFMWS 01One-, Two-, and Three-Dimensional, Multigroup, Discrete-Ordinates Transport Code System.
DCTDOSAbstractC00520 IBMPC 00Neutron and Gamma-Ray Penetration in Composite Duct Systems.
DDXCODESAbstractC00583 FM380 00One-, Two- and Three-Dimensional Transport Codes Using Multigroup Double-Differential Form Cross Sections.
DEMON & DEMON RAbstractC00181 I3675 00Demonstration Monte Carlo Code System in Slab Geometry.
DETAN 95AbstractP00361 MNYCP 00Code System to Calculate Spectrum-Averaged Cross Sections and Detector Responses in Neutron Spectra.
DIAMANT2AbstractC00414 PC386 00Multigroup Two-Dimensional Discrete Ordinates Transport Code System for Triangular Geometry, Release 2.0.
DISKTRANAbstractC00533 CYXMP 00Dose Calculations at Detectors from the End of a Cylinder Using DOT IV Scalar Flux Data.
DISKTRANAbstractC00533 I3033 00Dose Calculations at Detectors from the End of a Cylinder Using DOT IV Scalar Flux Data.
DKRAbstractC00323 CY000 00A Radioactivity and Dose Rate Calculation Code System for Fusion Reactors.
DLSAbstractC00264 C6600 00Two-Dimensional Shielding Calculational System with Diffusion Theory and Line-of-Sight Method.
DOORS 3.2AAbstractC00650 MFMWS 04One, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System.
DRAGON3.05DAbstractC00647 MNYWS 03Lattice Cell Code System.
DTF-INDIAAbstractC00458 I0370 00Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering.
DTF-IVAbstractC00042 C6600 00Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering.
DTF-IV MODIFIEDAbstractC00042 I0370 00Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering.
DTF-TRACAAbstractC00412 U1100 00Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering.
DTKAbstractC00223 I3675 00One-Dimensional Multigroup Neutron Transport Code System.
EASY-QAD 2.0.1AbstractC00744 PC586 02A Visualization Code System for Gamma and Neutron Shielding Calculations.
ELFAbstractC00167 I0360 00Monte Carlo Neutron Transport Code System for Cylinders and Spheres.
EMPIRE-IIAbstractP00497 PC586 01Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections.
ESPAbstractC00193 I0360 00General Purpose Monte Carlo Neutron Transport Code System.
EXIFON2.0AbstractP00305 IPCXT 01A Model for Statistical Multistep Direct and Multistep Compound Reactions.
EXTREMEAbstractC00440 I3033 00Two-Dimensional Discrete-Ordinates Code System with Exponential Expansion of Spatial Variables.
FASTER IIIAbstractC00168 U1108 00Monte Carlo Neutron and Photon Transport Code System in Complex Geometries.
FASTER-IIIAbstractC00168 I3675 00Monte Carlo Neutron and Photon Transport Code System in Complex Geometries.
FDKRAbstractC00541 I4381 00Radioactivity and Dose Rate Calculation Code for Fission, Fusion and Hybrid Reactors.
FEM-2DAbstractC00260 C6600 00Two-Dimensional Diffusion Theory Code System Based on the Method of Finite Elements.
FEMBAbstractC00340 B6700 00A Two-Dimensional Diffusion Theory Finite Element Program.
FEMRZAbstractC00342 F2307 00A Finite-Element Method Two-Dimensional Multigroup Neutron Transport Code System, (r,z) Geometry.
FLUKA05-PRE-LIBAbstractD00260 PCX86 00FLUKA05 Multi-Group, Multi-Purpose Nuclear Data Library, Neutrons, Photons, Charged Particles.
FLYSPEC-SHORTSAbstractP00196 C7600 00Neutron Unfolding Code System for Reducing Proton-Recoil Pulse-Height Obtained with NE-213 Liquid Scintillator.
FOCUSAbstractC00390 I3033 00Adjoint Monte Carlo Neutron Transport Code System.
FPZDAbstractC00603 PC386 00Code System for Multigroup Neutron Diffusion/Depletion Calculations.
FURNACEAbstractC00615 C0740 00Code System for Neutronic Calculations in Three Dimension Toroidal Geometry.
GBANISNAbstractC00628 IRISC 00One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering with the GroupBand Option.
GENP-2AbstractC00575 ALLMF 00Generalized Perturbation Theory Code System.
GIRAFFEAbstractP00304 I3033 00General Isotope Release Analysis For Failed Elements.
GRENADEAbstractC00516 C1787 00Green's Function Nodal Algorithm for the Diffusion Equation.
GRENADEAbstractC00516 D0780 00Green's Function Nodal Algorithm for the Diffusion Equation.
GROUPSTRUCTURESAbstractD00274 MNYCP 00GROUPSTRUCTURES, VITAMIN-J, XMAS, ECCO-33, ECCO2000 Standard Group Structures
GUI2QAD-3DAbstractC00697 PC586 01A Graphical User Interface for QAD-CGPIC, a Point Kernal Code for Neutron and Gamma-Ray Shielding Calculations in Complex Geometry.
HAMAbstractC00267 U1108 00Monte Carlo Multigroup Neutron and Photon High Altitude Transport Code System.
HEPROWAbstractC00799 MNYCP 00Unfolding of Pulse Height Spectra Using Bayes Theorem and Maximum Entropy Method.
HEXAB-3DAbstractC00593 I0370 00Three-Dimensional Few-Group Coarse Mesh Diffusion Code for Neutron Physics Calculation of Reactor Core in Hexagonal Geometry.
INAPAbstractC00235 U1108 00Improved Neutron Activation Prediction Code Systems.
INDRAAbstractC00303 I0360 00A Modular System for Calculating the Neutronics and Photonics Characteristics of a Fusion Reactor Blanket.
JN-METD 2&1AbstractC00208 I0370 00Neutron Transport Code System with Isotropic Scattering, Bare Slabs and Homogeneous Slabs (JN Method 1), Multilayer Slabs (JN Method 2).
KAMCCOAbstractC00325 I0370 00Three-Dimensional Time Dependent Monte Carlo Code System for Fast Neutron Physics Problems.
KAOS-VAbstractP00306 CY000 00An Evaluation Tool For Neutron Kerma Factors and Other Nuclear Responses.
KAP-VIAbstractC00094 U1108 00Kernel Integration Code System in Complex Geometry.
KDLIBEAbstractC00124 I3675 00Kernel-Diffusion Shielding Analysis System.
KIMAbstractC00376 I3033 00A Two-Dimensional Monte Carlo Code System for Linear Neutron Transport Calculations.
KORIGENAbstractC00457 I3033 00A Modification of the Isotope Generation and Depletion Code System ORIGEN.
LAHET 2.8AbstractC00696 MFMWS 00Code System for High Energy Particle Transport Calculations.
LASERAbstractC00344 I0360 00A One-Dimensional, Neutron-Thermalization, Lattice-Cell Program Based on MUFT and THERMOS.
LEOPARDAbstractC00343 C0000 00A Spectrum-Dependent Non-Spatial Fuel Depletion Code System.
LEOPARDAbstractC00343 IBMPC 00A Spectrum-Dependent Non-Spatial Fuel Depletion Code System.
LG-HAbstractC00087 I7090 00Ray Analysis Cylindrical Duct Kernel Code for Neutrons and Gamma Rays.
MADONNAAbstractC00425 I0370 00Two-dimensional Neutron Streaming Coupled Removal-Diffusion-Albedo-Transport Code System.
MAGIKAbstractC00359 I0360 00A Monte Carlo Code System for Computing Induced Residual Activation Dose Rates.
MAPAbstractC00150 I3675 00Kernel Integration Code System in Complex Geometry with Special Application to Surface Sources Determined by Discrete Ordinates Calculations.
MARC-PNAbstractC00311 D8810 00A Neutron Diffusion Code System with Spherical Harmonics Option.
MCNP6.2-EXE
810
AbstractC00850 MNYCP 01Monte Carlo N-Particle Transport Code System.
MCNP-DSP
810
AbstractC00699 MNYCP 00Monte Carlo N-Particle Transport Code System with Digital Signal Processing based on MCNP4A.
MCNP-DSP-EXE
810
AbstractC00699 MNYCP 01Monte 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.
MCNPX-POLIMI-EXE
810
AbstractC00791 MNYCP 01Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities.
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.
MEDUSA-PIJAbstractC00349 F2307 00One-Dimensional Lagrangian Code for Plasma Hydrodynamic Analysis of a Fusion Pellet Driven by Ion Beams.
MERCURE 4-82AbstractC00142 I3033 00Three-Dimensional Code System for Integrating Multigroup Line-of-Sight Attenuation Kernels by Monte Carlo Techniques.
MKENO-DARAbstractC00513 FM380 00Direct Angular Representation Monte Carlo Code for Criticality Safety Analysis
MMCRAbstractC00441 FM200 00Multigroup Monte Carlo Neutron and Photon Transport Code.
MOCAAbstractC00590 IPCAT 00Monte Carlo Criticality Code System for Hexagonal Geometries.
MOCUPAbstractP00365 DALPU 00MCNP/ORIGEN Coupling Utility Programs.
MOMENT IAbstractC00188 U1108 00Moments Method Neutron Transport Code System.
MONK 6.3
FEDC
AbstractC00393 I3033 00A General Purpose Monte Carlo Neutronics Code System.
MONTEBURNS 2.0AbstractP00455 MNYCP 02Automated, Multi-Step Monte Carlo Burnup Code System.
MORSE-ALBAbstractC00394 FM200 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-ANSI STD.AbstractC00127 I3675 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-BAbstractC00368 I0370 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-CAbstractC00431 C7600 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-CGAbstractC00203 C0000 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-CGAbstractC00203 CY000 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-CGAbstractC00203 D0VAX 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-CGAbstractC00203 I0360 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-CGAbstractC00203 U0000 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-CGAAbstractC00474 ALLCP 03Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSEC-SP2AbstractP00142 H6000 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-CVAbstractC00535 HM280 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-EAbstractC00258 I0360 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-EMPAbstractC00588 IBMPC 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-HAbstractC00471 I3081 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-LAbstractC00261 C6600 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MORSE-SGCAbstractC00277 C7600 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
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.
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.
MVP-GMVP IIAbstractC00739 MNYCP 00General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods.
NAAPROAbstractC00722 PC586 00Neutron Activation Analysis PRognosis and Optimization Code System.
NACTAbstractC00502 U1100 00Screening Program for Neutron Activation Products.
NAPAbstractC00101 I7090 00Multigroup Time-Dependent Neutron Activation Prediction Code.
NITRANAbstractC00582 FM380 00Neutron Transport Code System Based On Anisotropic Scattering.
NMTC/JAMAbstractC00717 PC586 00High Energy Particle Transport Code System.
NRNAbstractC00054 C6600 00Multigroup Removal-Diffusion Code System for Planes, Cylinders and Spheres.
NUCCONAbstractC00439 S7800 00A Code System for Calculation of Time-Dependent Nuclide Concentrations, Activity, Gamma-Ray Dose Rate and Biological Hazard Potential of Fusion Reactor Materials Due to Neutron Irradiation.
NUCWIZAbstractP00616 PCX86 00NucWiz
NUFACEAbstractP00284 CYXMP 00An Interface Code For The Calculation of Nuclear Responses.
NX1-NX2AbstractP00310 D0VAX 00Code System to Calculate Excitation Functions for (n,charged particle) Reactions.
O5RAbstractC00017 I3675 00A General-Purpose Monte Carlo Neutron Transport Code System.
O6RAbstractC00128 I3675 00A General-Purpose Monte Carlo Transport Code System.
OMEGAAbstractC00433 BESM6 00Monte Carlo Criticality Code System.
ONETRANAbstractC00266 C7600 00A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System.
ONETRANAbstractC00266 CY000 00A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System.
ONETRANAbstractC00266 I3033 00A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System.
OOSIIAbstractC00324 C0000 00Calculation of Isotropic Scattering by Particles for One-Dimensional and Three-Dimensional Transport in Slabs by Invariant Imbedding, Orders-of-Scattering Method, Including Check Calculations by Integral Transport Theory and Monte Carlo.
ORIGEN2.2AbstractC00371 ALLCP 03Isotope Generation and Depletion Code - Matrix Exponential Method.
ORIGEN-JENDL32AbstractC00703 MNYWS 00Isotope Generation and Depletion Code - Matrix Exponential Method.
ORIP_XXIAbstractC00731 PC586 02Computer Programs for Isotope Transmutation Simulations.
ORPHEE VIAbstractC00159 I3675 00Kernel Integration Code System - Attenuation of Fast Neutrons in Cylindrical Layers of Water and Dense Material.
OZMAAbstractC00406 I0370 00Calculation of Resonance Reaction Rates in Reactor Lattices Using Resonance Profile Tabulations.
PALLAS-1D(VII)AbstractC00380 FM380 00Multigroup Time-Independent Neutron Transport Code System for Plane or Spherical Geometry.
PALLAS-2DCY-FXAbstractC00391 FM380 00Multigroup Time-Independent Neutron Transport Code System for Plane or Spherical Geometry.
PATCH-7AbstractC00243 C0074 00Three-Dimensional Kernel Integration Code-Explicit Single Scattering Option.
PEQAG-2AbstractP00293 IPCAT 00A Pre-equilibrium Computer Code With Gamma Emission.
PIGGAbstractC00138 C3600 00A Multigroup One-Dimensional P-1 Radiation Transport Code System.
PREMORAbstractC00369 I0360 00A Point Reactor Exposure Code System for Survey Nuclear Analysis of Power Plant Performance.
PRIMEDANA-2AbstractC00490 I3081 00Collapses Multigroup Cross Sections and Obtains Reaction Parameters by Solving Transport or Diffusion Equations.
PROBAbstractC00287 I0370 00Multigroup One-Dimensional Transport Code System, Collision Probability Method.
PSU-LEOPARD/RBIAbstractC00563 IBMPC 01A Spectrum Dependent Non-Spatial Depletion Code.
QADAbstractC00048 I0360 00Kernel Integration Code System.
QAD-BSAAbstractC00346 C0000 00Kernel Integration Code System.
QAD-CGGP-AAbstractC00645 MNYCP 00Kernel Integration Code System.
QAD-P5AbstractC00048 C6400 00Kernel Integration Code System.
QAD-UEAbstractC00448 H6000 00Kernel Integration Code System.
RACCAbstractC00388 CY000 00A Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems.
RACCAbstractC00388 I3033 00A Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems.
RACC-PULSEAbstractC00639 MNYWS 00RACC Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems Modified for Pulsed/Intermittent Activation Analysis.
RADHEAT-V4AbstractC00300 FM380 00A Code System To Generate Multigroup Constants and Analyze Radiation Transport for Shielding Safety Evaluation.
RAFFLE/2AbstractC00279 C0176 00A General Purpose Monte Carlo Code System for Neutron Transport with Mixed Zone Geometry Option.
RAFFLE/2 MOD 2AbstractC00279 I0360 00A General Purpose Monte Carlo Code System for Neutron Transport with Mixed Zone Geometry Option.
RAIDAbstractC00083 I7090 00Monte Carlo Multibend Duct Shielding Code.
RASC-2DAbstractC00318 I0370 00Two-Dimensional Removal Diffusion Code Reactor Shielding Design Code System.
REACTORSHIELDING-NMSAbstractM00014 MNYCP 00REACTORSHIELDING-NMS, Reactor Shielding for Nuclear Engineers.
REDIFFUSIONAbstractC00347 I0360 00One-Dimensional Neutron Removal-Diffusion and Gamma-Ray Kernel Integration or Diffusion Theory Calculator.
REFIT-2009AbstractC00775 PCX86 00Multilevel Resonance Parameter Least Square Fit of Neutron Transmission, Capture, Fission & Self Indication Data.
RETRANSAbstractC00669 SUN05 00Code System For Calculating Reactivity Transients In a LWR.
REX2-87AbstractP00290 D8810 00A Code For Calculating Self-Shielded Multigroup Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data Files.
RFUNCAbstractP00312 D0VAX 00Code System to Analyze Differential Scattering Data.
RGENDFAbstractP00239 C0170 00Format Translation from NJOY GENDF Format to ENDF/B-V and Other Formats.
RHEINAbstractC00585 I3090 00Reactor Code System for Neutron Physics Calculation.
RICANTAbstractC00569 D8810 00A Computer Code for 2-D Transport Calculations in x-y Geometry Using the Interface Current Method.
RMET21AbstractC00597 D0VAX 00Detailed Space and Energy Treatment of Neutron Resonances for Homogeneous Mixtures and Cylinderized Reactor Cells.
ROCKWELL-RSDMAbstractM00017 MNYCP 00Reactor Shielding Design Manual by Rockwell T. III.
RSYSTAbstractC00269 I0360 00Integrated Modular Code System for Shielding and Reactor Physics Calculations.
SABINE-3AbstractC00121 C7600 00Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry.
SABINE-3AbstractC00121 I0370 00Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry.
SABINE-3AbstractC00121 U1106 00Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry.
SABINE-PCAbstractC00121 IBMPC 00Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry.
SAM-CEAbstractC00187 C6600 00Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations.
SAM-CEAbstractC00187 I0360 00Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations.
SAM-CEPAbstractC00192 C6600 00Monte Carlo Code System Correlated to the Simultaneous Solution of Multiple, Perturbed, Time-Dependent Neutron Transport Problems in Complex Three-Dimensional Geometry.
SAMSYAbstractC00315 C0073 00A One-Dimensional Multilayer Multigroup Neutron Removal-Diffusion and Gamma-Ray Point Kernel Calculator.
SAND-IIAbstractC00112 MNYCP 03Neutron Flux Spectra Determination by Multiple Foil Activation Method.
SAND-II-SNLAbstractP00345 SUN04 00Neutron Flux Spectra Determination by Multiple Foil Activation Method.
SANDORAbstractC00364 C7600 00Isotope Generation and Depletion Code Matrix Exponential Method.
SAP N-GAbstractC00092 I7094 00Neutron and Gamma-Ray Albedo Model Scatter Shield Analysis Code System.
SC2N3NAbstractP00309 D0VAX 00Systematics of (n,2n) and (n,3n) Cross Sections.
SCAP-82AbstractC00418 C7600 00Single Scatter, Albedo Scatter, or Point Kernel Analysis Code System in Complex Geometry.
SCAT-2AbstractP00294 MNYCP 03Code System for Calculating Total and Elastic Scattering Cross Sections Based on an Optical Model of the Spherical Nucleus.
SCORE-4AbstractC00234 I0370 00Two-Dimensional Multigroup Removal-Diffusion Shielding Code System.
SELFS-3AbstractP00551 C6600 00Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-II.
SENSITAbstractC00405 C7600 00One-Dimensional, Multigroup Cross Section and Design Sensitivity and Uncertainty Analysis Code System - Generalized Perturbation Theory.
SERA-1C1AbstractC00729 MNYCP 01Simulation Environment for Radiotherapy Applications.
SERPENTAbstractC00757 MNYWS 00Continuous Energy Monte Carlo Reactor Physics Burnup Calculation Code.
SHADOKAbstractC00216 C6600 00Transport Code Systems, P1 Scattering in Infinite Cylindrical and Spherical Geometries by Polynomial Approximation.
SHREDIAbstractC00284 I0360 00Multigroup Two-Dimensional (x-y, r-o geometry) Neutron Removal-Diffusion (Spinney Method) Shielding Code System.
SIXTUS-3AbstractC00609 MFMWS 00Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry.
SKYSHINE-KSUAbstractC00646 IBMPC 03Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air.
SLDNAbstractC00221 A1000 00Code System for Shielding Calculations by the Method of Invariant Imbedding.
SLDNAbstractC00221 F2307 00Code System for Shielding Calculations by the Method of Invariant Imbedding.
SLDNAbstractC00221 FM200 00Code System for Shielding Calculations by the Method of Invariant Imbedding.
SLDNAbstractC00221 GE625 00Code System for Shielding Calculations by the Method of Invariant Imbedding.
SLDNAbstractC00221 I0360 00Code System for Shielding Calculations by the Method of Invariant Imbedding.
SMAUG-13AbstractC00194 C6600 00Calculation of Neutron and Prompt Gamma-Ray Doses Resulting from an Atmospheric Nuclear Detonation.
SNAP-3DAbstractC00434 MNYCP 01Multigroup Complex Geometry Neutron Diffusion Code System.
SNEXAbstractC00353 C0000 00A One-Dimensional Single Group Discrete Ordinates Transport Code System.
SNOWAbstractC00282 I0360 00Two-Dimensional Discrete Ordinates Multigroup Transport Code System in Plane and Cylindrical Geometry with Isotropic and Anisotropic Scattering.
SOURCES-4CAbstractC00661 MNYCP 04Code System for Calculating (alpha,n), Spontaneous Fission, and Delayed Neutron Sources and Spectra.
SPACETRAN 1;2;3AbstractC00120 I3675 00Dose Calculations at Detectors at Various Distances from the Surface of a Cylinder.
SPECTER-ANLAbstractP00263 D0VAX 00Neutron Damage Calculations for Materials Irradiations.
SPECTRAAbstractC00108 C0000 00Determination of Neutron Spectra from Activation.
SPECTRAAbstractC00108 C0073 00Determination of Neutron Spectra from Activation.
SPECTRAAbstractC00108 C3600 00Determination of Neutron Spectra from Activation.
STRAINTAbstractC00259 I0360 00One-Dimensional Multigroup Neutron Transport Discrete Ordinates Code System.
SURFAbstractC00102 I3675 00Conical and Plane Surface Single Scattering Code.
SUSDAbstractC00501 HM150 00Cross Section Sensitivity and Uncertainty Analysis Including Secondary Neutron Energy and Angular Distributions.
SUSDAbstractC00501 I3090 00Cross Section Sensitivity and Uncertainty Analysis Including Secondary Neutron Energy and Angular Distributions.
SUSD3DAbstractC00695 MNYCP 01Multi-Dimensional, Discrete-Ordinates Based Cross Section Sensitivity and Uncertainty Analysis Code System.
SWANAbstractC00248 C0000 00Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics.
SWANAbstractC00248 CY000 00Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics.
SWANAbstractC00248 I0360 00Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics.
SWANLAKEAbstractC00204 C6600 00Cross Section Sensitivity Analysis Code System for One-Dimensional Discrete Ordinates Calculations.
SWANLAKEAbstractC00204 I3033 00Cross Section Sensitivity Analysis Code System for One-Dimensional Discrete Ordinates Calculations.
TALYS-1.2AbstractP00548 PC586 01Nuclear Model Code System for Analysis and Prediction of Nuclear Reactions and Generation of Nuclear Data.
TART2016AbstractC00638 MNYCP 08Coupled 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.
TEMPEST-2AbstractP00558 I0360 00Thermalization Program for Neutron Spectra and MultiGroup Cross-Sections.
TESSAbstractC00215 C3600 00Multigroup Discrete Ordinates Code System for Slab and Spherical Geometries.
THIDA-2AbstractC00410 FM380 00Code System for the Calculation of Transmutation, Activation, Decay Heat and Dose Rate in Fusion Reactors.
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.
TNG1AbstractP00298 D6220 00A Multistep Statistical Model Based on the Hauser-Feshbach Theory For The Evaluation Of Neutron Data.
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 Code System for the Calculation of Reactivity and Kinetic Parameters by One-Dimensional Neutron Transport Perturbation Theory.
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.
TRANZITAbstractC00172 C7600 00Multigroup Time-Dependent Discrete Ordinates Radiation Transport Code System in (rho,z) Cylindrical Geometry.
TRAXAbstractP00280 C0720 00A Program For Optics of Curved Crystal Neutron Spectrometers.
TRD-3AbstractC00362 I3033 00Two-Dimensional Removal-Diffusion Neutron Shielding Code System.
TREEDEAbstractC00326 C0000 00Monte Carlo Neutron Transport Code System Based on the Track Rotation Estimator.
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.
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 00Code System for Coupled Neutron, Photon, Electron, Positron, 3-D, Time Dependent, Monte-Carlo, Transport Calculations.
TWOTRANAbstractC00195 C6600 00Two-Dimensional Multigroup Discrete Ordinates Transport C System in (x,y), (r,theta), and (r,z) Geometries.
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.
UMG 3.3AbstractP00529 PC586 00Unfolding with Maxed and Gravel.
UNIFY-ECNAbstractP00288 C0170 00A Program to Calculate Fast Neutron Data for Structural Materials.
VALE 1.1AbstractC00613 IRISC 01A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries.
VALE 1.1AbstractC00613 PC386 01A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries.
VCSAbstractC00262 I0360 00Coupled Discrete Ordinates-Adjoint Monte Carlo Calculation of Radiation Protection Factors in Vehicles.
VENTURE-PCAbstractC00654 PC586 02A Reactor Analysis Code System.
VIEWCXSAbstractP00514 PC586 00Interactive Graphic User Interface to View Neutron and Gamma-Ray Interaction Cross Sections.
VIM 5.1AbstractC00754 MNYWS 01Continuous Energy Neutron and Gamma-ray Transport Code System.
VSOP94AbstractC00670 MNYWS 00Computer Code System for Reactor Physics and Fuel Cycle Simulation.
WIMKAL-88AbstractD00193 MNYCP 0069 Energy Group, Neutron Cross Section Library For Thermal Reactor Calculations in WIMSD Format.
XSDRNAbstractC00123 C0073 00Multigroup One-Dimensional Discrete Ordinates Spectral Averaging N Transport Code System.
XSDRNAbstractC00123 I0360 00Multigroup One-Dimensional Discrete Ordinates Spectral Averaging N Transport 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.