| Packages with Keyword: BURNUP |
| Package Name | Abstract | RSICC Tapelist | Title |
| 1DB-2DB-3DB | Abstract | C00741 PC586 00 | One-Dimensional Diffusion Code System for Nuclear Reactor. |
| 3DDT | Abstract | C00605 C6600 00 | Multigroup Diffusion Code System for Use in Fast Reactor Analysis. |
| ABLEIT-TRANS | Abstract | P00247 C0175 00 | Error Propagation Analysis for Burnup Calculation. |
ARC 11.2892
FEDC | Abstract | C00824 MNYCP 02 | Code System for Analysis of Nuclear Reactors. |
| BISON 1.5 | Abstract | C00464 HM200 00 | One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System. |
| BISON-C | Abstract | C00659 MNYWS 00 | One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System. |
| BOLD VENTURE IV | Abstract | C00459 I3033 00 | A Reactor Analysis Code System. |
| BOXER | Abstract | C00766 MNYWS 00 | Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations |
| BUCORST | Abstract | P00339 PC386 00 | A Code to Prepare Burnup-Dependent Multigroup Nuclear Reactor Source Terms. |
| CANDULIB-AECL | Abstract | D00210 MNYCP 00 | Burnup-Dependent ORIGEN-S Cross Section Libraries for CANDU Reactor Fuel Characterization. |
| CONDOR-3 | Abstract | C00811 I0370 00 | Two-Dimensional Reactor Program with Local and Spectrum Dependent Burnup. |
| DCHAIN-SP2001 | Abstract | C00712 MNYWS 01 | Code System for Analyzing Decay and Build-up Characteristics of Spallation Products. |
| DEPLETOR | Abstract | P00523 MNYCP 00 | Code System to Provide Depletion Capability to the U.S. NRC PARCS Code |
| DRAGON3.05D | Abstract | C00647 MNYWS 03 | Lattice Cell Code System. |
ERANOS 2.0
OECD | Abstract | C00745 MNYWS 00 | Modular Code and Data System for Fast Reactor Neutronics Analyses |
| FEMAXI 6 VER.1 | Abstract | P00536 IBMPC 00 | Code System for Light Water Reactor Fuel Analysis. |
| FPZD | Abstract | C00603 PC386 00 | Code System for Multigroup Neutron Diffusion/Depletion Calculations. |
| LASER | Abstract | C00344 I0360 00 | A One-Dimensional, Neutron-Thermalization, Lattice-Cell Program Based on MUFT and THERMOS. |
| LEOPARD | Abstract | C00343 C0000 00 | A Spectrum-Dependent Non-Spatial Fuel Depletion Code System. |
| LEOPARD | Abstract | C00343 IBMPC 00 | A Spectrum-Dependent Non-Spatial Fuel Depletion Code System. |
| MARIA SYSTEM | Abstract | P00359 D6000 00 | Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations. |
| 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. |
| ORIP_XXI | Abstract | C00731 PC586 02 | Computer Programs for Isotope Transmutation Simulations. |
| ORLIBJ32 | Abstract | D00255 MNYCP 00 | ORIGEN2 Libraries Based on JENDL-3.2. |
| PREMOR | Abstract | C00369 I0360 00 | A Point Reactor Exposure Code System for Survey Nuclear Analysis of Power Plant Performance. |
| PSU-LEOPARD/RBI | Abstract | C00563 IBMPC 01 | A Spectrum Dependent Non-Spatial Depletion Code. |
| PWR-AXBUPRO-GKN | Abstract | D00209 MNYCP 00 | Measured Axial Burnup Profiles for NeckarWesthiem PWR Reactors. |
| PWR-AXBUPRO-SNL | Abstract | D00201 MNYCP 00 | Axial Burnup Profile Database for Pressurized Water Reactors. |
| RAPID | Abstract | C00797 PCX86 00 | RAdial Power and Burnup Prediction by Following Fissile Isotope Distribution in the Pellet. |
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. |
| REBUS-PC 1.4 | Abstract | C00708 PC586 00 | Code System for Analysis of Research Reactor Fuel Cycles. |
| RICECCC | Abstract | C00348 I0360 00 | A Reactor Nuclide Inventory Code for Calculating Actinides and Fission Products. |
| SERPENT2.2.1 | Abstract | C00872 MNYWS 01 | Continuous Energy Monte Carlo Reactor Physics Burnup Calculation Code. |
| SRAC95 | Abstract | C00716 MNYWS 00 | Thermal Reactor Code System for Reactor Design and Analysis. |
| SWAT | Abstract | C00714 MNYCP 01 | Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2. |
| TRIGAP | Abstract | C00600 IBMPC 00 | A Computer Code for TRIGA Type Reactors. |
| TRIGLAV | Abstract | P00495 PC586 00 | Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor. |
| 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. |