| Packages starting with S |
| Package Name | Abstract | RSICC Tapelist | Title |
| S1CALC | Abstract | P00134 I0360 00 | A Multigroup Thermal Neutron Scattering Law Data Generator for Hydrogen and Deuterium. |
| S3 | Abstract | C00322 C6600 00 | Kernel Integration Code System--Multigroup Gamma-Ray Scattering. |
| S3 | Abstract | C00322 DVX11 00 | Kernel Integration Code System--Multigroup Gamma-Ray Scattering. |
| S3 | Abstract | C00322 IBMPC 00 | Kernel Integration Code System--Multigroup Gamma-Ray Scattering. |
| SABINE-3 | Abstract | C00121 C7600 00 | Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry. |
| SABINE-3 | Abstract | C00121 I0370 00 | Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry. |
| SABINE-3 | Abstract | C00121 U1106 00 | Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry. |
| SABINE-PC | Abstract | C00121 IBMPC 00 | Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry. |
| SACALC3 | Abstract | C00802 PCX86 00 | Calculates the Average Solid Angle Subtended by a Volume. |
| SACHET | Abstract | C00571 D8810 00 | A Computer Program To Evaluate The Dynamic Fission Product Inventories in the Multiple Compartment System of PWR's. |
| SAEROSA | Abstract | P00573 MNYCP 00 | Single-Species Aerosol Coagulation and Deposition with Arbitrary Size Resolution. |
| SAFE-D/SAFE-R | Abstract | P00496 MNYCP 00 | Code System for the Analysis of Component Failure Data with a Compound Statistical Model. |
| SAHYB-2 | Abstract | C00820 I0360 00 | Solution of Ordinary Differential Equation with User-Supplied Subroutine |
| SAIL | Abstract | D00057 I0360 00 | 23 Neutron, 17 Gamma-Ray Group ALBEDO DATA for Concrete and Steel, Based on DOT 1-1/2-D Calculations using DLC-31/FEWG1 Data. |
| SAILOR | Abstract | D00076 PC386 01 | Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. |
| SAIPS | Abstract | P00203 E1040 00 | Information Processing System for Calculating Neutron Spectra from Measured Reaction Rates. |
| SAIPS-PC | Abstract | P00295 IBMPC 00 | Information Processing System for Calculating Neutron Spectra from Measured Reaction Rates. |
| SALE3D | Abstract | P00443 CY000 00 | ICEd-ALE Treatment of 3-D Fluid Flow. |
| SAM-CE | Abstract | C00187 C6600 00 | Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations. |
| SAM-CE | Abstract | C00187 I0360 00 | Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations. |
| SAM-CEP | Abstract | C00192 C6600 00 | Monte Carlo Code System Correlated to the Simultaneous Solution of Multiple, Perturbed, Time-Dependent Neutron Transport Problems in Complex Three-Dimensional Geometry. |
| SAMCR | Abstract | P00487 U1100 00 | Code System for 2-D Elastodynamic Fracture Analysis. |
| SAMMY 8.1.0 | Abstract | P00158 MNYCP 13 | Code System for Multilevel R-Matrix Fits to Neutron and Charged-Particle Cross-Section Data Using Bayes' Equations. |
| SAMPO80 | Abstract | P00204 DGNOV 00 | Gamma-Ray Spectrum Analysis Method for Minicomputers. |
| SAMPO-LRC | Abstract | P00186 C6600 00 | Gamma-Ray Spectrum Analysis Code. |
| SAMSY | Abstract | C00315 C0073 00 | A One-Dimensional Multilayer Multigroup Neutron Removal-Diffusion and Gamma-Ray Point Kernel Calculator. |
| SAND-II | Abstract | C00112 MNYCP 03 | Neutron Flux Spectra Determination by Multiple Foil Activation Method. |
| SAND-II-SNL | Abstract | P00345 SUN04 00 | Neutron Flux Spectra Determination by Multiple Foil Activation Method. |
| SANDOR | Abstract | C00364 C7600 00 | Isotope Generation and Depletion Code Matrix Exponential Method. |
| SANDYL | Abstract | C00361 C0000 00 | A Monte Carlo Three-Dimensional Code System for Calculating Combined Photon-Electron Transport in Complex Systems. |
| SAP N-G | Abstract | C00092 I7094 00 | Neutron and Gamma-Ray Albedo Model Scatter Shield Analysis Code System. |
| SAPHIRE 8.0.9 | Abstract | P00608 PCX86 00 | Systems Analysis Programs for Hands-On Integrated Reliability Evaluations. |
| SARA 4.16 | Abstract | P00484 IBMPC 00 | System Analysis and Risk Assessment System. |
| SATURN | Abstract | P00057 I3675 00 | P1 or Transport Corrected Multigroup Neutron Cross Section Data Processor. |
| SC2N3N | Abstract | P00309 D0VAX 00 | Systematics of (n,2n) and (n,3n) Cross Sections. |
SCALE 6.3.1-EXE
810 | Abstract | C00860 MNYCP 03 | A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design |
| SCAMPI | Abstract | P00352 MNYWS 01 | Collection of Codes for Manipulating Multigroup Cross Section Libraries in AMPX Format. |
| SCANS | Abstract | P00029 I3675 00 | Spectra Calculation from Activated Nuclide Sets. |
| SCANS 1A | Abstract | P00373 PC386 01 | Shipping Cask Design Review Analysis. |
| SCAP-82 | Abstract | C00418 C7600 00 | Single Scatter, Albedo Scatter, or Point Kernel Analysis Code System in Complex Geometry. |
| SCAT-2 | Abstract | P00294 MNYCP 03 | Code System for Calculating Total and Elastic Scattering Cross Sections Based on an Optical Model of the Spherical Nucleus. |
SCDAP/RELAP5/MOD3.3-EXE
810 | Abstract | P00581 MNYCP 01 | A Best-Estimate Transient Simulation of Light Water Reactor Coolant Systems During a Severe Accident. |
SCEPTRE 1.1
FEDC | Abstract | C00807 PCX86 00 | Sandia Computational Engine for Particle Transport for Radiation Effects. |
SCEPTRE 1.7
FEDC | Abstract | C00826 PCX86 01 | Sandia Computational Engine for Particle Transport for Radiation Effects. |
| SCINFUL | Abstract | P00267 CY0MP 00 | Scintillator Full Response to Neutron Detection. |
| SCINFUL | Abstract | P00267 D8600 00 | Scintillator Full Response to Neutron Detection. |
| SCIP V1.1 | Abstract | C00749 PCX86 00 | Radioactive Surface Contamination Investigation Program. |
| SCOPE | Abstract | P00210 I3033 00 | Computer Code System for Shipping Cask Optimization and Parametric Evaluation. |
| SCORCH-B2 | Abstract | P00601 I0370 00 | BWR Core Heating During LOCA. |
| SCORE-4 | Abstract | C00234 I0370 00 | Two-Dimensional Multigroup Removal-Diffusion Shielding Code System. |
| SCORE-EVET | Abstract | P00442 C7600 00 | Code System for Three-Dimensional Hydraulic Reactor Core Analysis. |
| SCRELA | Abstract | P00408 SUN05 00 | Code System for Supercritical Water Cooled Reactor LOCA Analysis. |
| SDC | Abstract | C00060 I3675 00 | Kernel Integration Shield Design Code for Radioactive Fuel Handling Facilities. |
| SECA | Abstract | P00104 I0360 00 | Evaluator of Angular Bounds for a Two-Dimensional Symmetric Gaussian Quadrature Set. |
| SEDONE | Abstract | C00345 I0360 00 | A Simulator of Tidal Transient Hydrodynamic Sediment Concentrations Conditions in Controlled Rivers and Estuaries. |
| SEISIM1 | Abstract | P00453 C7600 00 | Code System for Seismic Probabilistic Risk Assessment. |
| SELFS-3 | Abstract | P00551 C6600 00 | Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-II. |
| SENPRO | Abstract | D00045 I3691 02 | Compilation of Multigroup Sensitivity Profiles in SENPRO Format for Fast Reactor Core and Shield Benchmarks and Thermal Reactor Benchmarks. |
| SENSIT | Abstract | C00405 C7600 00 | One-Dimensional, Multigroup Cross Section and Design Sensitivity and Uncertainty Analysis Code System - Generalized Perturbation Theory. |
| SERA-1C1 | Abstract | C00729 MNYCP 01 | Simulation Environment for Radiotherapy Applications. |
| SERPENT2.2.1 | Abstract | C00872 MNYWS 01 | Continuous Energy Monte Carlo Reactor Physics Burnup Calculation Code. |
| SESOIL | Abstract | C00629 IBMPC 03 | Code System to Calculate One-Dimensional Vertical Transport for the Unsaturated Soil Zone. |
| SETS | Abstract | P00380 CDCMF 00 | Set Equation Transformation System. |
| SFACTOR | Abstract | C00310 I0360 00 | Dose Equivalent to a Target Organ Calculator. |
| SFAK | Abstract | C00437 I3033 00 | Code System for Calculation of the Self-Absorption of Unscattered Gamma Radiation from Fuel Assemblies. |
SFHA
USSO | Abstract | P00413 IBMPC 00 | Code System for Spent Fuel Heating Analysis. |
| SHADOK | Abstract | C00216 C6600 00 | Transport Code Systems, P1 Scattering in Infinite Cylindrical and Spherical Geometries by Polynomial Approximation. |
| SHADRAC(G-30) | Abstract | C00084 I7090 00 | Kernel Integration Code - Shield Heating and Dose Rate Calculation in Complex Geometry. |
| SHAMSI | Abstract | D00135 I3033 00 | 48 Group Cross-Section Library for Fusion Nucleonics Analysis. |
| SHARDA | Abstract | C00521 C0740 00 | Sample Heat, Activity, Reactivity, and Dose Analysis for Safety Analysis of Irradiations in a Research Reactor. |
SHC
USSO | Abstract | P00493 CY000 00 | Seismic/Hazard Characterization in the Eastern U.S. |
| SHIELD | Abstract | C00667 MNYCP 01 | Monte-Carlo Code System to Simulate Interaction of High Energy Hadrons with Complex Macroscopic Targets. |
| SHIELDOSE | Abstract | C00379 ALLMF 00 | Code System for Space Shielding Radiation Dose Calculations. |
| SHIELDOSE-PC | Abstract | C00379 IBMPC 00 | Code System for Space Shielding Radiation Dose Calculations. |
| SHREDI | Abstract | C00284 I0360 00 | Multigroup Two-Dimensional (x-y, r-o geometry) Neutron Removal-Diffusion (Spinney Method) Shielding Code System. |
| SIGMA II | Abstract | C00118 C6000 00 | Space Radiation Dose Analysis Within Complex Configurations. |
| SIGMA II | Abstract | C00118 PC486 00 | Space Radiation Dose Analysis Within Complex Configurations. |
| SIGMA-A | Abstract | D00139 ALLMF 00 | Photon Interaction and Absorption Cross Sections. |
| SIGMA-A | Abstract | D00139 IBMPC 00 | Photon Interaction and Absorption Cross Sections. |
| SIGPI | Abstract | P00475 D0785 00 | Fault Tree Cut Set System Performance. |
SIMMER II
USSO | Abstract | C00691 MFMWS 00 | Code System for Two-Dinensional Sn-Neutronics and Fluid Dynamics. |
| SINBAD SEARCH TOOL | Abstract | P00580 MNYCP 00 | SINBAD Search Tool |
| SINBAD-2022 | Abstract | D00237 MNYCP 06 | Shielding Integral Benchmark Archive and Database. |
| SIOB | Abstract | P00139 I0360 00 | Calculation of Least-Squares Shape Fitting Several Neutron Transmission Measurements Using the Breit-Wigner Multilevel Formula. |
| SIR-3 | Abstract | P00055 C6400 00 | Sievert's Integral Routine-Computer Evaluation. |
| SIR-3 | Abstract | P00055 I3675 00 | Sievert's Integral Routine-Computer Evaluation. |
| SIXTUS-3 | Abstract | C00609 MFMWS 00 | Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry. |
| SKETCH-N 1.0 | Abstract | C00808 MNYCP 00 | Solve Neutron Diffusion Equations of Steady-State and Kinetics Problems. |
| SKEWGAUS | Abstract | P00089 I0360 00 | Skewed-Gaussian Line Peak Fitting Code - Multichannel Analyzer (MCA) Spectra - Ge(Li) and Semiconductor Detectors. |
| SKYDATA-KSU | Abstract | D00188 IBMPC 00 | Parameters for Approximate Neutron and Gamma-Ray Skyshine Response Functions and Ground Correction Factors. |
| SKYIII-PC | Abstract | C00289 IBMPC 01 | Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air. |
| SKYPORT | Abstract | D00093 IBMPC 00 | Skyshine Importance Functions for Neutrons and Gamma Rays. |
| SKYSHINE-III | Abstract | C00289 D0VAX 00 | Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air. |
| SKYSHINE-KSU | Abstract | C00646 IBMPC 03 | Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air. |
| SLAROM | Abstract | P00244 FM380 00 | A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors. |
| SLDN | Abstract | C00221 A1000 00 | Code System for Shielding Calculations by the Method of Invariant Imbedding. |
| SLDN | Abstract | C00221 F2307 00 | Code System for Shielding Calculations by the Method of Invariant Imbedding. |
| SLDN | Abstract | C00221 FM200 00 | Code System for Shielding Calculations by the Method of Invariant Imbedding. |
| SLDN | Abstract | C00221 GE625 00 | Code System for Shielding Calculations by the Method of Invariant Imbedding. |
| SLDN | Abstract | C00221 I0360 00 | Code System for Shielding Calculations by the Method of Invariant Imbedding. |
| SLIDERULE 1.0 | Abstract | C00704 PC586 01 | Nuclear Criticality Slide Rule. |
| SMACS | Abstract | P00396 C7600 01 | Probabilistic Seismic Analysis Code System. |
| SMAFS | Abstract | P00547 PC586 00 | Steady-State Analysis Model for Advanced Fuel Cycle Schemes. |
| SMART | Abstract | C00602 ALLCP 00 | Code System for Calculating Early Offsite Consequences from Nuclear Reactor Accidents. |
| SMART/MANYCASK | Abstract | C00482 FM200 00 | A Program for Calculating Radiation Dose Rates. |
| SMAUG-13 | Abstract | C00194 C6600 00 | Calculation of Neutron and Prompt Gamma-Ray Doses Resulting from an Atmospheric Nuclear Detonation. |
| SMOG | Abstract | P00216 I3033 00 | Code System for Neutron Cross Section Evaluation (Optical Method). |
| SNAKE | Abstract | P00135 I0360 00 | A Solid Angle Calculational System. |
| SNAP-3D | Abstract | C00434 MNYCP 01 | Multigroup Complex Geometry Neutron Diffusion Code System. |
| SNEX | Abstract | C00353 C0000 00 | A One-Dimensional Single Group Discrete Ordinates Transport Code System. |
| SNLRML | Abstract | D00178 ALLCP 00 | Recommended Dosimetry Cross Section Compendium. |
| SNOW | Abstract | C00282 I0360 00 | Two-Dimensional Discrete Ordinates Multigroup Transport Code System in Plane and Cylindrical Geometry with Isotropic and Anisotropic Scattering. |
| SOFIP | Abstract | C00358 I3033 00 | Evaluator of Space Radiation Environment Encountered by Geocentric Satellites. |
| SOFIRE-2 | Abstract | P00570 I0370 00 | Containment Temperature and Pressure During Na Pool Fire, 1-Cell or 2 Cell. |
| SOLA-DF | Abstract | P00454 C7600 00 | Code System to Calculate Transient 2-Dimensional 2-Phase Flow. |
| SOLA-LOOP | Abstract | P00464 C7600 00 | Nonequilibrium, Drift-Flux Code System for Two-Phase Flow Network Analysis |
| SOLTRAN | Abstract | C00763 PCX86 00 | Solving Multi-Dimensional Simplified P2 Transport and Diffusion Problems of Hexagonal Geometry in Fast Reactors. |
| SOPHIA | Abstract | C00857 MNYCP 00 | A Lagrangian-based computational fluid dynamics code for nuclear thermal hydraulics and safety applications. |
| SORA | Abstract | P00174 I0360 00 | A Code System for Storage and Retrieval of Data from Radionuclide Analyses. |
| SOSUM | Abstract | C00109 I3675 00 | Multigroup Beta and Gamma-Ray Energy Sources from Activities. |
| SOURCES-4C | Abstract | C00661 MNYCP 04 | Code System for Calculating (alpha,n), Spontaneous Fission, and Delayed Neutron Sources and Spectra. |
| SPACETRAN 1;2;3 | Abstract | C00120 I3675 00 | Dose Calculations at Detectors at Various Distances from the Surface of a Cylinder. |
| SPAR | Abstract | C00228 C6600 00 | Calculation of Stopping Powers and Ranges for Muons, Charged Pions, Protons and Heavy Ions. |
| SPAR | Abstract | C00228 I0360 00 | Calculation of Stopping Powers and Ranges for Muons, Charged Pions, Protons and Heavy Ions. |
| SPARES | Abstract | C00148 I3675 00 | Space Radiation Environment and Shielding Code System. |
| SPEC-4 | Abstract | P00099 I0360 00 | Calculated Recoil Proton Energy Distributions from Monoenergetic and Continuous Spectrum Neutrons. |
| SPECTER | Abstract | P00023 I3565 00 | Calculation of Energy Distribution of Nuclear Reaction Products. |
| SPECTER-ANL | Abstract | P00263 D0VAX 00 | Neutron Damage Calculations for Materials Irradiations. |
| SPECTRA | Abstract | C00108 C0000 00 | Determination of Neutron Spectra from Activation. |
| SPECTRA | Abstract | C00108 C0073 00 | Determination of Neutron Spectra from Activation. |
| SPECTRA | Abstract | C00108 C3600 00 | Determination of Neutron Spectra from Activation. |
| SPECTRANS-2 | Abstract | P00071 ICL00 00 | Neutron Spectrum Library Generation. |
| SPEEDI | Abstract | C00507 FM180 00 | Code System for Real-Time Prediction of Radiation Dose to the Public Due to an Accidental Release from a Nuclear Power Plant. |
| SPES | Abstract | P00602 I0370 00 | Fuel Cycle Optimization for LWR. |
| SPHINX | Abstract | P00129 C7600 00 | A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code System. |
| SPHINX | Abstract | P00129 I0360 00 | A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code System. |
SPIRT
USSO | Abstract | P00476 C7600 00 | Code System to Calculate Stress-Strains from Transient Pressures. |
SPIRT-NRC
USSO | Abstract | P00198 I3033 01 | Computerized Mathematical Models of Spray Washout of Airborne Contaminants (Radioactivity) in Containment Vessels. |
| SPOOR | Abstract | C00278 C7600 00 | Monte Carlo Simulation of the Turbulent Transport of Airborne Contaminants. |
| SPOT1 | Abstract | C00460 I3033 00 | Shielding Problem Code Based on Methods of Ono and Tsuruo. |
| SPUNIT | Abstract | P00266 D8600 00 | Spectrum Unfolding Using Information Theory. |
SQUIRT VER2
USSO | Abstract | P00583 PCX86 00 | Code System to Predict Leakage Rate and Area of Crack Opening for Cracked Pipes in Nuclear Power Plants. |
| SRAC95 | Abstract | C00716 MNYWS 00 | Thermal Reactor Code System for Reactor Design and Analysis. |
| SRNA-2K5 | Abstract | C00789 PCX86 00 | Proton Transport Simulation by Monte Carlo Techniques. |
SRVAL
USSO | Abstract | P00467 I3033 00 | Stock-Recruitment Model Validation Code System. |
SSC-L V3.3
USSO | Abstract | P00400 I3090 00 | Transient Response in LMFBR System. |
| STABA,STAGT,STEGT,STIG,STIGMA | Abstract | P00575 MNYCP 00 | Stress Analysis of Dragon HTR Graphite Structure. |
| STACY | Abstract | C00859 PCX86 00 | Source Term Analysis Code System. |
STANDARDS 5.0.1
RUGA | Abstract | C00873 PCX86 01 | Storage Transportation and Disposal Analysis Resource and Data System. |
| STAPREF | Abstract | P00498 PC586 00 | Code System to Calculate Nuclear Reaction Cross Sections by Evaporation Model. |
| STAPRE-H95 | Abstract | P00325 MNYCP 01 | Code System to Calculate Energy-Averaged Cross Sections of Particle Induced Nuclear Reactions. |
| STAR CODES | Abstract | P00330 IBMPC 00 | Code System for Calculating Stopping-Power and Range Tables for Electrons, Protons, and Helium Ions. |
| STAY'SL | Abstract | P00113 DP010 00 | Least Squares Dosimetry Unfolding Code System. |
| STAYSL PNNL | Abstract | P00589 PCX86 00 | STAYSL PNNL Suite of Software Tools. |
| STERNO | Abstract | C00057 C0000 00 | Two Dimensional Gamma-Ray Heating Kernel Integration Code. |
| STEX II | Abstract | M00010 MNYCP 00 | International Steam Explosion Experimental Data Base. |
| STOPOW88 | Abstract | C00790 MNYCP 00 | Stopping Power of Fast Ions in Matter. |
| STORM | Abstract | C00067 I7090 00 | Solar Flare Radiation Hazard to Earth Orbiting Vehicles. |
| STORM-ISRAEL | Abstract | D00015 I0360 01 | Evaluated Photon Interaction Library, ENDF/B File 23 Format. |
| STRADE | Abstract | P00252 I3081 00 | Stratified Random Design. |
| STRAGL | Abstract | C00201 C6600 00 | Calculation of Energy Loss Straggling of Heavy Charged Particles. |
| STRAINT | Abstract | C00259 I0360 00 | One-Dimensional Multigroup Neutron Transport Discrete Ordinates Code System. |
| STREAM | Abstract | C00321 C7600 00 | A Three-Dimensional Cylindrical-Geometry Monte Carlo Ray Tracing Code for Computing Light Transmission. |
| SUBDOSA-II | Abstract | C00270 U1100 00 | Calculation of External Gamma-Ray and Beta-Ray Doses from Accidental Atmospheric Releases of Radionuclides. |
| SUGGEL | Abstract | P00508 MNYWS 00 | Program Suggesting the Orbital Angular Momentum of a Neutron Resonance From the Magnitude Of Its Neutron Width. |
| SULSA | Abstract | M00015 MNYCP 00 | A Solution for the Neutron Spectrum Unfolding Problem Without Using Input Spectrum (Report Only). |
| SUPERDAN-PC | Abstract | P00282 IBMPC 00 | Calculates Dancoff Factor of Spheres, Cylinders and Slabs. |
| SUPERTOG III M2 | Abstract | P00013 I3691 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-4 | Abstract | P00013 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-JR. | Abstract | P00115 F2307 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-JR. | Abstract | P00115 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-LTT | Abstract | P00228 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SURF | Abstract | C00102 I3675 00 | Conical and Plane Surface Single Scattering Code. |
| SUSD | Abstract | C00501 HM150 00 | Cross Section Sensitivity and Uncertainty Analysis Including Secondary Neutron Energy and Angular Distributions. |
| SUSD | Abstract | C00501 I3090 00 | Cross Section Sensitivity and Uncertainty Analysis Including Secondary Neutron Energy and Angular Distributions. |
| SUSD3D | Abstract | C00695 MNYCP 01 | Multi-Dimensional, Discrete-Ordinates Based Cross Section Sensitivity and Uncertainty Analysis Code System. |
| SWAN | Abstract | C00248 C0000 00 | Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics. |
| SWAN | Abstract | C00248 CY000 00 | Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics. |
| SWAN | Abstract | C00248 I0360 00 | Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics. |
| SWANLAKE | Abstract | C00204 C6600 00 | Cross Section Sensitivity Analysis Code System for One-Dimensional Discrete Ordinates Calculations. |
| SWANLAKE | Abstract | C00204 I3033 00 | Cross Section Sensitivity Analysis Code System for One-Dimensional Discrete Ordinates Calculations. |
| SWAP-9 | Abstract | C00788 C0740 00 | 1-D Stress Analysis for Hydrostatic and Elastic Plastic Materials. |
| SWAT | Abstract | C00714 MNYCP 01 | Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2. |
| SWIFT | Abstract | C00679 C7600 00 | Code System to Calculate Waste-Isolation Flow and Transport. |
| SWIFT | Abstract | P00031 C6600 00 | Monte Carlo Neutron Spectra Unfolding Code. |
SWIFT2
USSO | Abstract | C00686 MNYCP 00 | Code System to Calculate Waste-Isolation Flow and Transport. |
| SWORD 7.0 | Abstract | C00767 MNYCP 07 | SoftWare for Optimization of Radiation Detectors. |
| SYVAC-D/2 | Abstract | C00690 D0VAX 00 | Code System For Risk Assessment From Underground Radioactive Waste Disposal In the United Kingdom. |