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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: FUEL CYCLE AND WASTE MANAGEMENT |
| Package Name | Abstract | RSICC Tapelist | Title |
| 1DB-2DB-3DB | Abstract | C00741 PC586 00 | One-Dimensional Diffusion Code System for Nuclear Reactor. |
| ALPHN | Abstract | C00612 IBMPC 00 | Code System for Calculating (alpha,n) Neutron Production in Canisters of High-Level Waste. |
| AT123D | Abstract | C00417 I0360 00 | Analytical Transient One-, Two-, and Three-Dimensional Simulation of Waste Transport in an Aquifer 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. |
| CARL 2.3 | Abstract | C00743 PC586 01 | Code System to Calculate Radiotoxicity, Activity, Dose and Decay Power Calculations for Spent Fuel. |
| CASK | Abstract | D00023 I3691 04 | 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis. |
| CASK-81 | Abstract | D00023 I0370 05 | 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis. |
| CASK-81 | Abstract | D00023 IBMPC 06 | 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis. |
| COBRA-EN | Abstract | P00507 MNYCP 01 | Thermal-Hydraulic Transient Analysis of Reactor Cores. |
| COBRA-SFS VERSION 6.0 | Abstract | P00614 MNYCP 02 | COBRA-SFS Thermal-Hydraulic Analysis of Multi-Assembly Spent Fuel Storage and Transportation Systems. |
DANESS V1.0
FEDC | Abstract | P00555 MNYCP 00 | Dynamic Analysis of Nuclear Energy System Strategies. |
| DEIS | Abstract | C00455 C6600 00 | Draft Environmental Impact Statement on Licensing Requirements for Land Disposal of Radioactive Waste. |
| DIFMOD | Abstract | C00572 I3083 00 | A Computer Program To Calculate The Leaching of Radionuclides and the Corrosion of Cemented Waste Forms in Water or Brine. |
| DPCT | Abstract | C00580 CYXMP 00 | A Deterministic-Probabilistic Model For Contaminant Transport. |
| DUST-BNL | Abstract | C00634 PC386 00 | Disposal Unit Source Term by One-Dimensional, Transient, Finite-Difference, Subsurface Release and Transport of Contaminants. |
ERANOS 2.0
OECD | Abstract | C00745 MNYWS 00 | Modular Code and Data System for Fast Reactor Neutronics Analyses |
| FAMREC | Abstract | P00167 C7600 01 | Fuel Assembly Mechanical Response Code System. |
| FASTGRASS | Abstract | P00479 MNYCP 00 | Code System to Predict Fission Product Release in Ubase Fuels. |
| FCXSEC | Abstract | D00085 PC386 01 | 22 Neutron, 21 Gamma-Ray Group Cross Section Libraries in ANISN Format for Nuclear Fuel Cycle Shielding Calculations. |
| FEMAXI 6 VER.1 | Abstract | P00536 IBMPC 00 | Code System for Light Water Reactor Fuel Analysis. |
| FPIC | Abstract | C00028 I3675 00 | Fission Product Inventory Code. |
| FPZD | Abstract | C00603 PC386 00 | Code System for Multigroup Neutron Diffusion/Depletion Calculations. |
| FRANCO | Abstract | P00363 MNYCP 00 | Finite Element Fuel Rod Analysis Code System. |
| FUELSDATA | Abstract | P00446 C7600 00 | Code System to Model Verification Fuel Rod Data. |
| GAPCON-THERMAL | Abstract | P00499 C7600 00 | Code System to Calculate Fuel Steady State & Transient Behavior. |
| GRASS-SST | Abstract | P00489 MNYCP 00 | Code System to Predict Fission-Gas Release & Fuel Swelling. |
| INTRUDE-ANS | Abstract | C00539 D8810 00 | A Repository Intrusion Risk Evaluation Code. |
LAS CRUCES
USSO | Abstract | D00194 ALLCP 00 | Las Cruces Trench Site Database, Vadose Model. |
| 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. |
| MCB1C | Abstract | C00719 MNYWS 00 | Monte-Carlo Continuous Energy Burnup Code System. |
| MONTEBURNS 2.0 | Abstract | P00455 MNYCP 02 | Automated, Multi-Step Monte Carlo Burnup Code System. |
| MURE V2-SMURE | Abstract | C00764 MNYWS 01 | Serpent - MCNP Utility for Reactor Evolution. |
| MYRA | Abstract | C00056 C0000 00 | Calculation of Shipping Costs and Cask Designs for Irradiated Fuel Elements. |
| MYRA | Abstract | C00056 I7090 00 | Calculation of Shipping Costs and Cask Designs for Irradiated Fuel Elements. |
| NCSP-DAT | Abstract | M00002 MNYCP 01 | Nuclear Data in Support of the Nuclear Criticality Safety Program. |
| NUTRAN | Abstract | C00675 I0370 00 | Code System for Long-Term Repository Safety Analysis. |
| ORIGEN2.2 | Abstract | C00371 ALLCP 03 | Isotope Generation and Depletion Code - Matrix Exponential Method. |
| ORIGEN-JENDL32 | Abstract | C00703 MNYWS 00 | Isotope Generation and Depletion Code - Matrix Exponential Method. |
| PAGAN | Abstract | C00621 IBMPC 00 | Code System for Performance Assessment Ground-water Analysis for Low-level Nuclear Waste. |
| PART61 | Abstract | C00499 IBMPC 01 | Low-Level Radioactive Waste Impacts Analysis System. |
| 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. |
| QBF | Abstract | C00617 PC386 00 | Code System to Calculate Radiation Dose Rates Relative to Spent Fuel Shipping Casks. |
| RADSYS | Abstract | C00530 I3033 00 | Code System for Radioactivity Buildup and Radioactive Waste Generation Calculations. |
| RATAF | Abstract | C00681 IMFPC 01 | Code System for the Radioactive Liquid Tank Failure Study. |
| 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. |
| REPRISK PC 1.02 | Abstract | C00586 PC386 01 | Repository Risk Assessment Software for Personal Computers. |
| SACHET | Abstract | C00571 D8810 00 | A Computer Program To Evaluate The Dynamic Fission Product Inventories in the Multiple Compartment System of PWR's. |
| SCANS 1A | Abstract | P00373 PC386 01 | Shipping Cask Design Review Analysis. |
| 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. |
| SYVAC-D/2 | Abstract | C00690 D0VAX 00 | Code System For Risk Assessment From Underground Radioactive Waste Disposal In the United Kingdom. |
| TRIGLAV | Abstract | P00495 PC586 00 | Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor. |
| 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. |
| WREM-TOODEE2 | Abstract | P00469 ALLMF 00 | 2-D Time-Dependent Fuel Element, Thermal Analysis Code System. |