Online Catalog
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810 -- US DOE 10CFR810 Jurisdiction
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OECD -- Restricted/See Abstract
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Packages with Keyword: REACTOR PHYSICS
Package NameAbstractRSICC TapelistTitle
1DB-2DB-3DBAbstractC00741 PC586 00One-Dimensional Diffusion Code System for Nuclear Reactor.
AIREKMOD-RRAbstractP00588 D0VAX 00Reactivity Transients in Nuclear Research Reactors
AIREKMOD-RRAbstractP00588 PCX86 01Reactivity Transients in Nuclear Research Reactors
ANL-BPBAbstractM00004 MNYCP 00Argonne National Laboratory Code Center: Benchmark Problem Book.
ARC 11.2892
AbstractC00824 MNYCP 02Code System for Analysis of Nuclear Reactors.
AUS98AbstractC00519 MNYWS 01Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems.
BCGAbstractC00578 C0170 00A Code For Calculating Pointwise Neutron Spectra and Criticality in Fast Reactor Cells.
BOLD VENTURE IVAbstractC00459 I3033 00A Reactor Analysis Code System.
CITATION-LDI 2AbstractC00643 PC386 02Nuclear Reactor Core Analysis Code System.
COBRA-ENAbstractP00507 MNYCP 01Thermal-Hydraulic Transient Analysis of Reactor Cores.
DANCOFF3AbstractP00279 D8810 00Calculates Dancoff Correction.
DASQHEAbstractP00278 D8810 00Calculates Dancoff Corrections Factors.
DIF3D 11.2892
AbstractC00784 MNYCP 02Code System Using Variational Nodal Methods and Finite Difference Methods to Solve Neutron Diffusion and Transport Theory Problems.
DRAGON3.05DAbstractC00647 MNYWS 03Lattice Cell Code System.
EACRP-D2O-LATTICESAbstractD00264 MNYCP 00Compilation of Reactor Physics Measurements in HWRs Lattices.
AbstractC00745 MNYWS 00Modular Code and Data System for Fast Reactor Neutronics Analyses
FEDGROUPC86REV3AbstractP00194 MNYCP 01Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation.
FEMAXI 6 VER.1AbstractP00536 IBMPC 00Code System for Light Water Reactor Fuel Analysis.
FPZDAbstractC00603 PC386 00Code System for Multigroup Neutron Diffusion/Depletion Calculations.
FURNACEAbstractC00615 C0740 00Code System for Neutronic Calculations in Three Dimension Toroidal Geometry.
GENP-2AbstractC00575 ALLMF 00Generalized Perturbation Theory Code System.
GERESAbstractP00241 I0370 00A Code to Produce Cross-Section Libraries for ANISN Based on Heterogeneous Fast Reactor Cell Calculations Using MC2II Data.
GT2R2AbstractP00483 ALLMF 00Code System to Calculate Fuel Rod Thermal Performance.
HEXAB-3DAbstractC00593 I0370 00Three-Dimensional Few-Group Coarse Mesh Diffusion Code for Neutron Physics Calculation of Reactor Core in Hexagonal Geometry.
IRPHE-VENUS-RECYCLEAbstractD00263 MNYCP 00Plutonium Recycling Physics Project Critical Experiments.
LTCAbstractP00329 IBMPC 00LMR Transient Calculation Code System.
MARCOPOLOAbstractP00225 I0360 00Code System for Calculating the Radial and Axial Neutron Diffusion Coefficients in One-Group and Multigroup Theory.
MCRACAbstractC00562 IBMPC 00Multiple Cycle Reactor Analysis Code.
MICROX-2AbstractP00374 MNYCP 02Code System to Create Broad-Group Cross Sections with Resonance Interference and Self-Shielding from Fine-Group and Pointwise Cross Sections.
MIGROS3AbstractP00265 I0370 00A Code for the Generation of Group Constants for Reactor Calculations from Neutron Nuclear Data in KEDAK Format.
MINETAbstractP00490 CY000 00Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis.
MKENO-DARAbstractC00513 FM380 00Direct Angular Representation Monte Carlo Code for Criticality Safety Analysis
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.
NEACRP-H2O-LATTICESAbstractD00265 MNYCP 00Compilation of Reactor Physics Measurements in LWRs Lattices.
NESTLE 5.2.1AbstractC00641 MNYCP 04Code System to Solve the Few-Group Neutron Diffusion Equation Utilizing the Nodal Expansion Method (NEM) for Eigenvalue, Adjoint, and Fixed-Source
OMEGAAbstractC00433 BESM6 00Monte Carlo Criticality Code System.
PERSENT 11.2892
AbstractC00823 MNYCP 02Perturbation and Sensitivity Code for Assembly Homogenized Multi-group Transport Problems
AbstractC00822 MNYWS 01Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS 11.2892
AbstractC00822 MNYCP 02Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS3/VARIANT8AbstractC00653 MNYWS 01Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS-PC 1.4AbstractC00708 PC586 00Code System for Analysis of Research Reactor Fuel Cycles.
RHEINAbstractC00585 I3090 00Reactor Code System for Neutron Physics Calculation.
RMET21AbstractC00597 D0VAX 00Detailed Space and Energy Treatment of Neutron Resonances for Homogeneous Mixtures and Cylinderized Reactor Cells.
AbstractC00691 MFMWS 00Code System for Two-Dinensional Sn-Neutronics and Fluid Dynamics.
SLAROMAbstractP00244 FM380 00A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors.
SUPERDAN-PCAbstractP00282 IBMPC 00Calculates Dancoff Factor of Spheres, Cylinders and Slabs.
TRIGAPAbstractC00600 IBMPC 00A Computer Code for TRIGA Type Reactors.
TRISTAN-IJSAbstractP00537 IBMPC 00Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System.
VENTURE-PCAbstractC00654 PC586 02A Reactor Analysis Code System.
VIM 5.1AbstractC00754 MNYWS 01Continuous Energy Neutron and Gamma-ray Transport Code System.
VPI-NECMAbstractC00481 C0740 00Nuclear Engineering Computer Models for In-Core Fuel Management Analysis.
VPI-NECMAbstractC00481 D0VAX 00Nuclear Engineering Computer Models for In-Core Fuel Management Analysis.
VPI-NECMAbstractC00481 PC486 00Nuclear Engineering Computer Models for In-Core Fuel Management Analysis.
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.
WIMS-ANL 4.0AbstractC00698 MNYCP 00Deterministic Code System for Reactor Lattice Calculation.
WIMSD-5B.12AbstractC00656 MNYCP 02Deterministic Code System for Reactor Lattice Calculation
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.