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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 Keyword: WORKSTATION |
| Package Name | Abstract | RSICC Tapelist | Title |
| ANISN-ORNL | Abstract | C00254 MNYCP 02 | One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering. |
| BETA-S 6 | Abstract | C00657 MNYCP 01 | Code System to Calculate Multigroup Beta-Ray Spectra. |
| BISON-C | Abstract | C00659 MNYWS 00 | One-Dimensional Discrete Ordinate Transport and Burnup Calculation 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. |
| BSPRP2 | Abstract | P00372 IRISC 00 | Code System to Process DORT Boundary-Flux Files. |
| CALOR95 | Abstract | C00610 MNYWS 00 | Monte Carlo Code System for Design and Analysis of Calorimeter Systems, Spallation Neutron Source (SNS) Target Systems, etc. |
| CCRMN | Abstract | P00366 MNYCP 00 | Monte Carlo Simulation of the Coupled Transport of Electrons and Photons. |
| CGS 11.4 | Abstract | P00243 MFMWS 03 | Common Graphics System. |
| CHENDF 7.02 | Abstract | P00333 MNYCP 05 | Codes for Handling ENDF/B-V and ENDF/B-VI Data. |
| DANTSYS 3.0 | Abstract | C00547 MFMWS 01 | One-, Two-, and Three-Dimensional, Multigroup, Discrete-Ordinates Transport Code System. |
| DCHAIN 1.3 | Abstract | C00640 MNYCP 01 | Code System for Radioactive Decay and Reaction Chain Calculations. |
| DETAN 95 | Abstract | P00361 MNYCP 00 | Code System to Calculate Spectrum-Averaged Cross Sections and Detector Responses in Neutron Spectra. |
| DOORS 3.2A | Abstract | C00650 MFMWS 04 | One, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System. |
| DRAGON3.05D | Abstract | C00647 MNYWS 03 | Lattice Cell Code System. |
| EGS4 | Abstract | C00331 MNYCP 00 | Monte Carlo Simulation of the Coupled Transport of Electrons and Photons. |
| EXPRESS | Abstract | C00622 MNYCP 00 | Exact Preparedness Supporting System. |
| GBANISN | Abstract | C00628 IRISC 00 | One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering with the GroupBand Option. |
| GICX40 | Abstract | D00092 ALLCP 00 | Coupled 42-Neutron, 21-Gamma-Ray Group Cross Sections for 40 Elements in Group Independent Form for Fusion Reactor Calculations. |
| GMA | Abstract | P00367 MNYCP 00 | Code System for Calculation of Reactor Accident Consequences. |
| GNASH-FKK | Abstract | P00535 MNYCP 00 | Pre-equilibrium, Statistical Nuclear-Model Code System for Calculation Cross Sections and Emission Spectra. |
| GRESS 3.0 | Abstract | P00231 MFMWS 02 | Gradient Enhanced Software System. |
| HEATING 7.3 | Abstract | P00199 MNYCP 06 | Multidimensional, Finite-Difference Heat Conduction Analysis Code System. |
| LAHET 2.8 | Abstract | C00696 MFMWS 00 | Code System for High Energy Particle Transport Calculations. |
| LEPRICON | Abstract | P00277 I3033 01 | PWR Pressure Vessel Surveillance Dosimetry Analysis System. |
| LEPRICON | Abstract | P00277 IRISC 00 | PWR Pressure Vessel Surveillance Dosimetry Analysis System. |
| MARLOWE 15B | Abstract | P00137 MNYCP 08 | Computer Simulation of Atomic Collisions in Crystalline Solids. |
MCNP-DSP-EXE
810 | Abstract | C00699 MNYCP 01 | Monte Carlo N-Particle Transport Code System with Digital Signal Processing based on MCNP4A. |
| MOCUP | Abstract | P00365 DALPU 00 | MCNP/ORIGEN Coupling Utility Programs. |
| MONTEBURNS 2.0 | Abstract | P00455 MNYCP 02 | Automated, Multi-Step Monte Carlo Burnup Code System. |
| MORSE-CGA | Abstract | C00474 ALLCP 03 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MSM-SOURCE | Abstract | P00369 MNYCP 00 | Code System for Generation of Input Data for MCNP. |
| NESTLE 5.2.1 | Abstract | C00641 MNYCP 04 | Code System to Solve the Few-Group Neutron Diffusion Equation Utilizing the Nodal Expansion Method (NEM) for Eigenvalue, Adjoint, and Fixed-Source |
| NJOY91.119 | Abstract | P00171 MFMWS 04 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY94.61 | Abstract | P00355 MFMWS 03 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY97.0 | Abstract | P00368 MNYCP 00 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY99.0 | Abstract | P00480 MNYCP 00 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| PARET-ANL | Abstract | P00516 MNYCP 00 | Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores. |
| PENELOPE-MPI | Abstract | C00713 IBMSP 00 | Code System for Monte Carlo Simulation of Electron and Photon Transport. |
| PREPRO2019 | Abstract | P00351 MNYCP 10 | Pre-Processing Code System for Data in ENDF/B Format. |
| QAD-CGGP-A | Abstract | C00645 MNYCP 00 | Kernel Integration Code System. |
| RACC-PULSE | Abstract | C00639 MNYWS 00 | RACC Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems Modified for Pulsed/Intermittent Activation Analysis. |
| REAC*3 | Abstract | C00443 IBMPC 00 | Computer Code System for Activation and Transmutation. |
| REAC*3 | Abstract | C00443 MFMWS 00 | Computer Code System for Activation and Transmutation. |
REBUS 11.0 EXE_ONLY
FEDC | Abstract | C00822 MNYWS 01 | Code System for Analysis of Fast Reactor Fuel Cycles. |
REBUS 11.2892
FEDC | Abstract | C00822 MNYCP 02 | Code System for Analysis of Fast Reactor Fuel Cycles. |
| REBUS3/VARIANT8 | Abstract | C00653 MNYWS 01 | Code System for Analysis of Fast Reactor Fuel Cycles. |
| ROLAIDS-CPM | Abstract | P00353 SUN04 00 | Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method. |
| SAND-II-SNL | Abstract | P00345 SUN04 00 | Neutron Flux Spectra Determination by Multiple Foil Activation Method. |
| SCAMPI | Abstract | P00352 MNYWS 01 | Collection of Codes for Manipulating Multigroup Cross Section Libraries in AMPX Format. |
| SCRELA | Abstract | P00408 SUN05 00 | Code System for Supercritical Water Cooled Reactor LOCA Analysis. |
| SOURCES-4C | Abstract | C00661 MNYCP 04 | Code System for Calculating (alpha,n), Spontaneous Fission, and Delayed Neutron Sources and Spectra. |
| TART2022 | Abstract | C00638 MNYCP 09 | Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code System. |
| TIBSO | Abstract | C00512 MNYCP 00 | Code System to Calculate Production and Migration of Radionuclides in Nuclear Reactor Systems. |
| TRANSX 2.15 | Abstract | P00317 MFMWS 01 | Code system to produce neutron, photon, and particle transport tables for discrete-ordinates and diffusion codes from cross sections in MATXS format. |
| VALE 1.1 | Abstract | C00613 IRISC 01 | A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries. |
| VALE 1.1 | Abstract | C00613 PC386 01 | A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries. |
| VIM 5.1 | Abstract | C00754 MNYWS 01 | Continuous Energy Neutron and Gamma-ray Transport Code System. |
| WIMKAL-88 | Abstract | D00193 MNYCP 00 | 69 Energy Group, Neutron Cross Section Library For Thermal Reactor Calculations in WIMSD Format. |
| WIMSD-5B.12 | Abstract | C00656 MNYCP 02 | Deterministic Code System for Reactor Lattice Calculation |
| ZOTT99 | Abstract | P00272 ALLCP 02 | Zero-in On The Truth; Evaluation of Correlated Data Using Partitioned Least Squares. |