RSICC Catalog

Online Catalog

Click on Package Name to get detailed information.
Click on Abstract to read the package abstract.
Click on RSICC Tapelist to view list of files distributed with package.

Note: RESTRICTIONS APPLY TO SOME PACKAGES -

810 -- US DOE 10CFR810 Jurisdiction
FEDC -- US Government Agencies and Their Contractors Only
OECD -- Restricted/See Abstract
RUGA -- Restricted Use Government Authorized
USSO -- US Distribution Only
USUNV -- US Universities Only

Packages with Subject: REACTOR PHYSICS
Package Name	Abstract	RSICC Tapelist	Title
BOLD VENTURE IV	Abstract	C00459 I3033 00	A Reactor Analysis Code System.
BOT3P-5.3	Abstract	P00530 MNYCP 02	Code System for 2D and 3D Mesh Generation and Graphical Display of Geometry and Results for Radiation Transport Codes.
CANDULIB-AECL	Abstract	D00210 MNYCP 00	Burnup-Dependent ORIGEN-S Cross Section Libraries for CANDU Reactor Fuel Characterization.
CARMEN SYSTEM	Abstract	C00487 U1110 00	A Code System for Neutronics PWR Calculation by Diffusion Theory with Space-Dependent Feedback Effects.
CITATION-LDI 2	Abstract	C00643 PC386 02	Nuclear Reactor Core Analysis Code System.
COBRA-EN	Abstract	P00507 MNYCP 01	Thermal-Hydraulic Transient Analysis of Reactor Cores.
DANCOFF-MC	Abstract	P00509 MNYCP 00	Code System for Monte Carlo Calculation of Dancoff Factors in Irregular Geometries.
DIF3D 11.2892 FEDC	Abstract	C00784 MNYCP 02	Code System Using Variational Nodal Methods and Finite Difference Methods to Solve Neutron Diffusion and Transport Theory Problems.
EACRP-D2O-LATTICES	Abstract	D00264 MNYCP 00	Compilation of Reactor Physics Measurements in HWRs Lattices.
EMPIRE-II	Abstract	P00497 PC586 01	Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections.
ERANOS 2.0 OECD	Abstract	C00745 MNYWS 00	Modular Code and Data System for Fast Reactor Neutronics Analyses
GT2R2	Abstract	P00483 ALLMF 00	Code System to Calculate Fuel Rod Thermal Performance.
HATCHES-19	Abstract	D00206 PC586 02	Thermodynamic Database for Radiochemical Modelling.
HEXAB-3D	Abstract	C00593 I0370 00	Three-Dimensional Few-Group Coarse Mesh Diffusion Code for Neutron Physics Calculation of Reactor Core in Hexagonal Geometry.
IRPHE-VENUS-RECYCLE	Abstract	D00263 MNYCP 00	Plutonium Recycling Physics Project Critical Experiments.
JDL-REACTOR-KIN	Abstract	M00006 MNYCP 00	Nuclear Reactor Kinetics and Control.
MINET	Abstract	P00490 CY000 00	Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis.
MOSRA-LIGHT	Abstract	P00505 MNYWS 00	High-Speed Three-Dimensional Nodal Diffusion Code System.
NEACRP-H2O-LATTICES	Abstract	D00265 MNYCP 00	Compilation of Reactor Physics Measurements in LWRs Lattices.
NORMA	Abstract	P00471 PC586 00	Code System to Solve Burnup Dependent Neutron Diffusion Equations in Two and Three Dimensions.
NORMA-FP	Abstract	P00470 PC586 00	Code System to Perform Neutronic and Thermal-Hydraulic Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions.
PWR-AXBUPRO-SNL	Abstract	D00201 MNYCP 00	Axial Burnup Profile Database for Pressurized Water Reactors.
RETRAC	Abstract	C00635 D0VAX 00	Code System for the Analysis of Material Test Reactor (MTR) Cores.
RMET21	Abstract	C00597 D0VAX 00	Detailed Space and Energy Treatment of Neutron Resonances for Homogeneous Mixtures and Cylinderized Reactor Cells.
TDTORT	Abstract	C00709 MNYWS 00	Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System.
TEMPEST-2	Abstract	P00558 I0360 00	Thermalization Program for Neutron Spectra and MultiGroup Cross-Sections.
TEMPEST-BNW	Abstract	P00559 C7600 00	Transient 3-D Thermohydraulics for FBR.
THT	Abstract	C00480 I0360 00	Three-Dimensional Neutron Coarse Mesh Code System to Evaluate Average Bundle Fluxes and Power in Light Water Reactors.
THYDE-B1/MOD2	Abstract	P00553 FM200 00	Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
THYDE-P2	Abstract	P00554 FV100 00	Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
TPTRIA	Abstract	C00550 I3083 00	A Computer Program for the Reactivity and Kinetic Parameters for Two-Dimensional Triangular Geometry by Transport Perturbation Theory.
TRIGLAV	Abstract	P00495 PC586 00	Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor.
TRISTAN-IJS	Abstract	P00537 IBMPC 00	Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System.
VENTEASY	Abstract	C00776 PCX86 00	Criticality search for a desired Keffective by adjusting dimensions, nuclide concentrations, or buckling
VENTURE-PC	Abstract	C00654 PC586 02	A Reactor Analysis Code System.
WIMS-ANL 4.0	Abstract	C00698 MNYCP 00	Deterministic Code System for Reactor Lattice Calculation.
WIMSD-5B.12	Abstract	C00656 MNYCP 02	Deterministic Code System for Reactor Lattice Calculation

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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.