| Package Name |
Abstract |
RSICC Tapelist |
Title |
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1DB-2DB-3DB
|
Abstract
|
C00741 PC586 00 |
One-Dimensional Diffusion Code System for Nuclear Reactor. |
1DX |
Abstract
|
P00096 I0360 00 |
A One-Dimensional Diffusion Code System for Producing Energy Group Collapsed and Self-Shielded Cross Sections. |
1DX |
Abstract
|
P00096 U1108 00 |
A One-Dimensional Diffusion Code System for Producing Energy Group Collapsed and Self-Shielded Cross Sections. |
3DDT
|
Abstract
|
C00605 C6600 00 |
Multigroup Diffusion Code System for Use in Fast Reactor Analysis. |
ABAREX |
Abstract
|
P00248 MNYCP 01 |
Neutron Spherical Optical-Statistical Model Code System. |
ABBN-90 |
Abstract
|
D00182 MNYCP 00 |
Multigroup Constant Set for Calculation of Neutron and Photon Radiation Fields and Functionals, Including the CONSYST2 Program. |
ABLEIT-TRANS |
Abstract
|
P00247 C0175 00 |
Error Propagation Analysis for Burnup Calculation. |
ACAT |
Abstract
|
P00257 FM380 00 |
Monte Carlo Simulation of Atomic Collisions in Amorphous Targets in the Binary Collision Approximation. |
ACDOS3
|
Abstract
|
C00442 C7600 00 |
Calculation of Activities and Dose Rates Produced by Neutron Activation. |
ACFA
|
Abstract
|
C00478 I3033 00 |
A Versatile Activation Code for Coolant and Structural Materials. |
ACOH
|
Abstract
|
C00191 I3675 00 |
Aerojet COHORT Monte Carlo Code System. |
ACORNS |
Abstract
|
P00264 IBMPC 01 |
Analysis of Correlations Used in Neutron Spectrometry. |
ACRA-II
|
Abstract
|
C00213 I0360 00 |
Kernel Integration Code System for Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident. |
ACRA-TRIT
|
Abstract
|
C00283 I0360 00 |
The Tritium Version of ACRA-II, Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident. |
ACRO
|
Abstract
|
C00354 I0360 00 |
Calculation of Organ Dose from Acute or Chronic Inhalation and Ingestion of Radionuclides. |
ACT-ARA
|
Abstract
|
C00372 CYXMP 00 |
Code System for the Calculation of Changes in Radiological Source Terms with Time. |
ACTIV-PC |
Abstract
|
P00287 IBMPC 00 |
A Program to Process Gamma or X-ray Spectra. |
ACTIV87 |
Abstract
|
D00169 ALLCP 00 |
Fast Neutron Activation Cross Section File. |
ACTL82 |
Abstract
|
D00069 ALLCP 01 |
Evaluated Neutron Activation Cross-Section Library. |
ACTV-F/H |
Abstract
|
D00155 ALLCP 00 |
Neutron Activation Cross Section Library for Fusion Reactor Design. |
ACTV-FUS/INT |
Abstract
|
D00170 ALLCP 00 |
International Library of Neutron Activation Cross-Section Data for Fusion Reactor Application. |
ADASAGE |
Abstract
|
P00426 IBMPC 00 |
Ada Application Development System, Versions 4.02, 4.0 and 3.1. |
ADEFTA 4.0 |
Abstract
|
P00543 MNYCP 00 |
Atomic Densities for Transport Analysis Script. |
ADENA |
Abstract
|
P00190 C0000 00 |
Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra. |
ADENA |
Abstract
|
P00190 I3033 00 |
Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra. |
ADJMOM
|
Abstract
|
C00212 I3675 00 |
Adjoint Moments Method Gamma-Ray Transport Code System. |
ADLER III |
Abstract
|
P00058 I0360 00 |
A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters. |
ADO
|
Abstract
|
C00189 I3675 00 |
Aerojet Discrete Ordinates Calculational System. |
AGDATA |
Abstract
|
D00127 I0360 00 |
Two Agricultural Production Data Libraries (AGDATC and AGDATG) for Dose and Risk Assessment Models. |
AIR DATA |
Abstract
|
D00014 I0360 00 |
Sample Input to ANISN for Calculation of Neutron and Secondary Gamma-Ray Transport in Air. |
AIRBORNE
|
Abstract
|
C00263 I0360 00 |
Airborne Contaminants Dispersion Code. |
AIRDIF
|
Abstract
|
C00360 C6600 00 |
A Two-Dimensional Atmospheric Radiation Diffusion Code. |
AIRDOS-PC
|
Abstract
|
C00551 IBMPC 00 |
Clean Air Act Compliance Software for Personal Computers. See C00542/CAP-88. |
AIREM
|
Abstract
|
C00242 I3691 00 |
Calculation of Doses, Population Doses, and Ground Depositions Due to Atmospheric Emissions of Radionuclides. |
AIRFEWG |
Abstract
|
D00049 I0360 00 |
Results of ANISN Multigroup Calculations of Gamma-Ray, Neutron, and Secondary Gamma-Ray Transport in Infinite Homogeneous Air Using DLC-31/(DPL-1/FEWG1) Cross Sections. |
AIRGAMMA
|
Abstract
|
C00567 FM380 00 |
A Program For The Calculation Of External Exposure To Gamma Rays From A Radioactive Cloud. |
AIRSCAT
|
Abstract
|
C00341 DP010 00 |
Calculation of Dose Rate for Gamma-Rays Scattered in Air. |
AIRTRANS
|
Abstract
|
C00110 I3675 00 |
Monte Carlo Time and Energy-Dependent Three-Dimensional Radiation Transport Code. |
AISITE II
|
Abstract
|
C00286 I0360 00 |
Reactor Siting Code System. |
AKERN
|
Abstract
|
C00190 C0000 00 |
Aerojet Point Kernel Integration Calculational System. |
AKERN
|
Abstract
|
C00190 U1108 00 |
Aerojet Point Kernel Integration Calculational System. |
AKTIV
|
Abstract
|
C00339 I0360 00 |
An Evaluation of Activity, Afterheat and Biological Hazard Potential of Stainless Steel Structures in Fusion Reactor Blankets. |
ALARA 2.7.8
|
Abstract
|
C00723 MNYCP 00 |
Code System for Analytic and Laplacian Adaptive Radioactivity Analysis. |
ALARM-B2 |
Abstract
|
P00218 I0360 00 |
A Computer Code System for Analysis of a Large Break LOCA of a BWR. |
ALBEDO-DATA |
Abstract
|
D00224 MNYCP 00 |
KSU Neutron Albedo Data. |
ALBEMO
|
Abstract
|
C00268 C6600 00 |
Albedo Monte Carlo Code System. |
ALDOSE
|
Abstract
|
C00577 IBMPC 00 |
Dose Calculation for Alpha Disc Source. |
ALEPH-LIB-JEFF3.1 |
Abstract
|
D00230 MNYCP 00 |
ACE Format Neutron Cross Section Library based on JEFF3.1. |
ALGAM-97
|
Abstract
|
C00152 I3675 00 |
Monte Carlo Estimation of Internal Dose from Gamma-Ray Sources in a Phantom Man. |
ALICE-91 |
Abstract
|
P00146 MNYCP 02 |
Statistical Model Code System with Fission Competition. |
ALKASYS-PC
|
Abstract
|
C00558 IBMPC 00 |
A Computer Program For Studies of Rankine-Cycle Space Nuclear Power Systems. |
ALPHA-M |
Abstract
|
P00169 I0360 00 |
Least-Squares Resolution of Gamma-Ray Spectra in Environmental Samples. |
ALPHN
|
Abstract
|
C00612 IBMPC 00 |
Code System for Calculating (alpha,n) Neutron Production in Canisters of High-Level Waste. |
ALPS |
Abstract
|
P00144 F2307 00 |
Alpha Spectrum Analysis Code System. |
AMARA |
Abstract
|
P00079 I3675 00 |
Nuclear Data Adjustment Using Lagrange's Multipliers Method. |
AMC
|
Abstract
|
C00090 I3675 00 |
Monte Carlo Albedo Code for Neutron and Capture Gamma-Ray Distributions in Rectangular Concrete Ducts. |
AMPX 77 |
Abstract
|
P00315 CY000 00 |
Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
AMPX-77 |
Abstract
|
P00315 ALLMF 01 |
Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
AMPX01 |
Abstract
|
D00027 I3675 02 |
126-Group Coupled Neutron and Gamma-Ray Transport Cross-Section Data Generated by AMPX. |
AMUSE |
Abstract
|
P00028 C6600 00 |
Gamma-Ray Spectra Unfolding Code. |
ANA |
Abstract
|
P00356 IBMPC 00 |
Code System for Gamma-Ray Spectra Analyses. |
ANIPLO D50 |
Abstract
|
P00213 I0360 00 |
A Digital Computer Program for Plotting Results from Calculations with the Sn Computer Program ANISN. |
ANISN-ORNL
|
Abstract
|
C00254 MNYCP 02 |
One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering. |
ANISN-PC
|
Abstract
|
C00514 IBMPC 00 |
Multigroup One-Dimensional Discrete Ordinates Transport Code System with
Anisotropic Scattering. RSIC recommends CCC-650/DOORS3.2a for most applications. |
ANITA-2000
|
Abstract
|
C00693 MNYCP 00 |
Code System to Calculate Isotope Inventories from Neutron Irradiation for Fusion Applications. |
ANITA-4
|
Abstract
|
C00606 MNYCP 01 |
Analysis of Neutron Induced Transmutation and Activation. See ANITA-2000 (CCC-693). |
ANL-BPB
| Abstract
|
M00004 MNYCP 00 |
Argonne National Laboratory Code Center: Benchmark Problem Book. |
ANS643 |
Abstract
|
D00129 IBMPC 02 |
ANS-6.4.3 Geometric Progression Gamma-Ray Buildup Factor Coefficients. |
ANSIFT |
Abstract
|
P00077 C6600 00 |
ANSI Standard Fortran Sifting Program. |
ANSIFT |
Abstract
|
P00077 I0360 00 |
ANSI Standard Fortran Sifting Program. |
ANSL-V |
Abstract
|
D00154 ALLCP 01 |
ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies. |
ANTE 2
|
Abstract
|
C00131 I3675 00 |
Adjoint Monte Carlo Time-Dependent Neutron Transport Code in Combinatorial Geometry. |
APARNA-II
|
Abstract
|
C00296 I0360 00 |
Integral Transport Theory Code System Based on Discrete Ordinate Representation in Space and Direction-Slab Geometry. |
APPLE-2 |
Abstract
|
P00111 FM200 00 |
Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates. |
APPLE-2 |
Abstract
|
P00111 I3081 00 |
Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates. |
APSAI |
Abstract
|
P00065 I3691 00 |
Activity Calculations and Plotting of Neutron or Gamma-Ray Spectra Generated by Discrete Ordinates Code System ANISN. |
APUD 3.0
|
Abstract
|
C00637 IBMPC 00 |
Code System for Analyzing, Predicting Consequences of, and Guiding the Response to Nuclear Emergencies. |
ARC
|
Abstract
|
C00224 C6600 00 |
Aircraft Radiation Transport Code System, Crew Dose Calculation. |
ARCON96
|
Abstract
|
C00664 IBMPC 00 |
Code System to Calculate Atmospheric Relative Concentrations in Building Wakes. |
AREAC
|
Abstract
|
C00438 I3033 00 |
Radiological Emission Analysis Code System. |
AREAD |
Abstract
|
P00088 I0360 00 |
Input Data Processor for Transport Codes. |
ARMYL-G
|
Abstract
|
C00297 U1106 00 |
Calculation of Transmission Factors for Gamma Rays from Nuclear Explosions. |
ARMYL-N
|
Abstract
|
C00298 U1106 00 |
Calculation of Transmission Factors for Neutrons from Nuclear Explosions. |
ARRRG
|
Abstract
|
C00404 U1100 00 |
Calculation of Radiation Dose to Man from Radionuclides in the Environment. |
ASFIT-VARI
|
Abstract
|
C00336 H0000 00 |
Gamma-Ray Transport Code System for One-Dimensional Finite Systems. |
ASFIT-VARI
|
Abstract
|
C00336 IBMPC 00 |
Gamma-Ray Transport Code System for One-Dimensional Finite Systems. |
ASOP
|
Abstract
|
C00126 IRISC 00 |
Multigroup One-Dimensional Discrete Ordinates Transport Code System for Shield Optimization. |
ASTROS
|
Abstract
|
C00073 I7090 00 |
Calculation of Primary and Secondary Proton Dose Rates in Spheres and Slabs of Tissue. |
AT123D
|
Abstract
|
C00417 I0360 00 |
Analytical Transient One-, Two-, and Three-Dimensional Simulation of Waste Transport in an Aquifer System. |
ATHENA_2D |
Abstract
|
P00431 MNYCP 00 |
Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum. |
ATM-TOX
|
Abstract
|
C00472 I3033 00 |
An Atmospheric Transport Model for Toxic Substances. |
ATTOW-KB
|
Abstract
|
C00132 I0370 00 |
Multigroup Two-Dimensional Removal-Diffusion (Spinney Method) Shielding Code System. |
AUS98
|
Abstract
|
C00519 MNYWS 01 |
Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems. |
AUTOJOM-JOMREAD |
Abstract
|
P00008 C6600 00 |
Computer Programs to Generate or Check Coefficients for Quadratic Equations Describing 3D Geometries. |
AXMIX |
Abstract
|
P00075 CYXMP 00 |
ANISN Cross Section Code System. |
AXMIX |
Abstract
|
P00075 IRISC 01 |
ANISN Cross Section Code System. |
AXMIX-PC |
Abstract
|
P00297 IBMPC 00 |
ANISN Cross Section Mixing Code System. |
BABEL |
Abstract
|
D00104 I3033 00 |
Multi-Purpose Neutron and Gamma-Ray Cross Section Library for Fast Reactor Shielding Design. |
BALTORO
|
Abstract
|
C00479 C6600 00 |
Code for Coupling of Monte Carlo and Discrete Ordinates Radiation Transport Calculations. |
BARC-35 |
Abstract
|
D00124 IBMMF 00 |
35-Group Neutron Cross Sections and Resonance Self-Shielding Factors Generated in ISOTXS and BRKOXS Format from ENDF/B-IV Using MINX. |
BASACF |
Abstract
|
P00285 IBMPC 00 |
Bayesian Approach to Spectrum Adjustment with Covariance Filter. |
BAYES |
Abstract
|
P00205 DP010 00 |
User's Guide for A General-Purpose Computer Code System for Fitting a Functional Form to Experimental Data. |
BCG
|
Abstract
|
C00578 C0170 00 |
A Code For Calculating Pointwise Neutron Spectra and Criticality in Fast Reactor Cells. |
BEACON MOD3 |
Abstract
|
P00402 CDCMF 00 |
Code System for Thermal-Hydraulic Analysis of Nuclear Reactor Containments. |
BEBC
|
Abstract
|
C00077 I7090 00 |
Electron Bremsstrahlung Penetration Code for Space Vehicles. |
BED
|
Abstract
|
C00078 I7090 00 |
Electron Penetration Code for Space Vehicles. |
BERMUDA
|
Abstract
|
C00616 FV260 03 |
Discrete Ordinates Code System for Shielding Analysis for Use with Fusion and Fission Reactors. |
BETA II
|
Abstract
|
C00117 C6600 00 |
Monte Carlo Bremsstrahlung and Electron Transport Analysis in Geometry. |
BETA II
|
Abstract
|
C00117 I0360 00 |
Monte Carlo Bremsstrahlung and Electron Transport Analysis in Geometry. |
BETA-S 6
|
Abstract
|
C00657 MNYCP 01 |
Code System to Calculate Multigroup Beta-Ray Spectra. |
BFR
USSO
|
Abstract
|
P00449 C0176 00 |
Code System for Common Cause Failure Data Analysis. |
BIGGI
|
Abstract
|
C00066 I3675 00 |
Numerical Gamma-Ray Transport Code for Plane or Spherical Multilayer Geometry, Versions 3P &4T. |
BIGGI
|
Abstract
|
C00066 U1108 00 |
Numerical Gamma-Ray Transport Code for Plane or Spherical Multilayer Geometry, Versions 3P and 4T. |
BISON 1.5
|
Abstract
|
C00464 HM200 00 |
A One-dimensional Discrete Ordinate Transport and Burnup Calculation Code. (Burnup of Isotopes and One-Dimensional Transport). |
BISON-C
|
Abstract
|
C00659 MNYWS 00 |
One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System. |
BLOCKAGE V2.5R |
Abstract
|
P00377 IBMPC 00 |
Code System to Calculate Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in a BWR. |
BLT-FEMWATER
USSO
|
Abstract
|
C00633 PC386 00 |
Code System to Solve for Release and Transport of Contaminants through Saturated/Unsaturated Media. |
BMC-MG
|
Abstract
|
C00291 C6600 00 |
Multigroup Monte Carlo Neutron and Gamma-Ray Shielding Code System for Plutonium. |
BOB-7 SERIES |
Abstract
|
P00084 F2306 00 |
Theory and Use of Gamma-Ray Spectrum Analysis Codes for Ge(Li) Detectors. |
BOLD VENTURE IV
|
Abstract
|
C00459 I3033 00 |
A Reactor Analysis Code System. |
BON |
Abstract
|
P00173 I0360 00 |
A Code System for Unfolding Multisphere Spectrometer Neutron Measurements. |
BOT3P-5.2 |
Abstract
|
P00530 MNYCP 01 |
Code System for 2D and 3D Mesh Generation and Graphical Display of Geometry and Results for Radation Transport Codes. |
BPPC
|
Abstract
|
C00076 I7090 00 |
Proton Penetration Codes for Space Vehicles. |
BREESE-II |
Abstract
|
P00143 I3033 00 |
Auxiliary Routines for Implementing the Albedo Option in the MORSE Monte Carlo Code System. |
BREMRAD
|
Abstract
|
C00031 I7090 00 |
External and Internal Bremsstrahlung Calculation Code. |
BRHGAM
|
Abstract
|
C00350 I3033 00 |
Monte Carlo Estimation of Absorbed Dose from X-Ray Sources in Phantom Man. |
BRMSTK |
Abstract
|
P00044 C6600 00 |
CSEWG Integral Data Testing Shielding Experiment Code System. |
BRMSTK |
Abstract
|
P00044 I3691 00 |
CSEWG Integral Data Testing Shielding Experiment Code System. |
BSPRP2 |
Abstract
|
P00372 IRISC 00 |
Code System to Process DORT Boundary-Flux Files. |
BUCORST |
Abstract
|
P00339 PC386 00 |
A Code to Prepare Burnup-Dependent Multigroup Nuclear Reactor Source Terms. |
BUGLE-80 |
Abstract
|
D00075 PC386 01 |
Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96. |
BUGLE-80 |
Abstract
|
D00075 IBMPC 02 |
Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96. |
BUGLE-80 |
Abstract
|
D00075 IBMPC 03 |
Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96. |
BUGLE-96 |
Abstract
|
D00185 ALLCP 00 |
Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. |
BULK_C-12
|
Abstract
|
C00738 PC586 00 |
Code System to Estimate Neutron and Photon Effective Dose Rates from Medium Energy Protons or Carbon Ions Through Concrete or Concrete/Iron. |
BURP-2
|
Abstract
|
C00237 C6600 00 |
Calculation of Buildup and Decay of Radioactive Fission Products. |
BUSH
|
Abstract
|
C00333 I0360 00 |
A Code to Calculate Radiation Doses Inside Buildings from Routine Releases of Radionuclides to the Atmosphere. |
BWR-LTAS
|
Abstract
|
C00485 I3033 01 |
Code System for Boiling Water Reactor Long-Term Accident Simulation. |
CAAC
|
Abstract
|
C00476 D0VAX 00 |
Code System for Implementation of Atmospheric Dispersion Assessment Required by the Clear Air Act. See CCC-542/CAP-88. |
CAAC
|
Abstract
|
C00476 I3033 00 |
Code System for Implementation of Atmospheric Dispersion Assessment Required by the Clear Air Act. See CCC-542/CAP-88. |
CACA-2
|
Abstract
|
C00302 I0360 00 |
Heavy Isotope and Fission-Product Concentration Calculation Code System. |
CAD |
Abstract
|
D00059 I0360 00 |
51 Neutron, 25 Gamma-Ray Group ALBEDO DATA Generated with DOT for Various Materials. |
CALENDF-2005 |
Abstract
|
P00539 MNYCP 00 |
Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations. |
CALKUX
|
Abstract
|
C00594 IBMPC 00 |
Code System to Calculate Exposure Transmission of Medical X-ray Beams Through Barrier Materials. |
CALOR95
|
Abstract
|
C00610 MNYWS 00 |
Monte Carlo Code System for Design and Analysis of Calorimeter Systems, Spallation Neutron Source (SNS) Target Systems, etc. |
CAMERA
|
Abstract
|
C00240 C0074 00 |
Radiation Transport Analysis Code System and the Computerized Man (CAM) Model. |
CAMERA
|
Abstract
|
C00240 IBMPC 01 |
Radiation Transport Analysis Code System and the Computerized Man (CAM) Model. |
CANDULIB-AECL |
Abstract
|
D00210 MNYCP 00 |
Burnup-Dependent ORIGEN-S Cross Section Libraries for CANDU Reactor Fuel Characterization. |
CAP-88
|
Abstract
|
C00542 D0VAX 00 |
Clean Air Act Assessment Package. |
CAP-88
|
Abstract
|
C00542 I3090 00 |
Clean Air Act Assessment Package. |
CAP88-PC
|
Abstract
|
C00542 IBMPC 01 |
Clean Air Act Assessment Package. |
CAPS-2
|
Abstract
|
C00074 CDCMF 00 |
Analysis of Structures for Fallout Radiation Shielding. |
CARL-2.2
|
Abstract
|
C00743 PC586 00 |
Code System to Calculate Radiotoxicity, Activity, Dose and Decay Power Calculations for Spent Fuel. |
CARMEN SYSTEM
|
Abstract
|
C00487 U1110 00 |
A Code System for Neutronics PWR Calculation by Diffusion Theory with Space-Dependent Feedback Effects. |
CARNAC
|
Abstract
|
C00238 I3691 00 |
Calculation of Flux and Neutron Spectra in the Case of Criticality Accident. |
CARP-82 |
Abstract
|
P00131 I3033 00 |
Multigroup Albedo Data Using DOT Angular Flux Results. |
CARSTEP
|
Abstract
|
C00024 I7090 00 |
Trajectory and Environment Code-Electron and Proton Fluxes Impinging on Spacecraft in Orbit. |
CASCADE
|
Abstract
|
C00176 C6600 00 |
Monte Carlo Simulation of the Transport of High Energy Electrons and Photons in Matter. |
CASCADE
|
Abstract
|
C00176 I0360 00 |
Monte Carlo Simulation of the Transport of High Energy Electrons and Photons in Matter. |
CASIM
|
Abstract
|
C00265 I0360 00 |
Monte Carlo Simulation of Transport of Hadron Cascades in Bulk Matter. |
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. |
CASKCODES |
Abstract
|
P00262 IBMPC 00 |
CAPSIZE, SCOPE, AND KWIKDOSE for Shipping Cask Optimization, Dose Calculation, Parameter Evaluation, and Shielding Requirements. |
CASTHY |
Abstract
|
P00316 F0230 00 |
Statistical Model Calculation for Neutron Cross Sections and Capture Gamma-Ray Spectra. |
CAVEAT
|
Abstract
|
C00169 I3675 00 |
General Purpose Monte Carlo Time-Dependent Radiation Transport Code in Complex Geometry. |
CCRMN |
Abstract
|
P00366 MNYCP 00 |
Monte Carlo Simulation of the Coupled Transport of Electrons and Photons. |
CDR
|
Abstract
|
C00182 C6600 00 |
A Constant Dose Range Code System, Using the LANL-NWEF Neutron-Gamma-Ray Air Flux Tape. |
CDR
|
Abstract
|
C00182 I0360 00 |
A Constant Dose Range Code System, Using the LANL-NWEF Neutron-Gamma-Ray Air Flux Tape. |
CEAR-PPU |
Abstract
|
P00528 PC586 00 |
Code System for Monte Carlo Simulation of Detector Pulse Pile Up. |
CECP-BWR |
Abstract
|
P00370 PC386 00 |
Estimating Boiling Water Reactor Decomissioning Costs. |
CECP-PWR |
Abstract
|
P00371 PC386 00 |
Estimating Pressurized Water Reactor Decomissioning Costs. |
CEM03.01 |
Abstract
|
P00532 MNYCP 00 |
Monte-Carlo Code System to Calculate Nuclear Reactions in the Framework of Improved Cascade-Exciton Model. |
CEMENT 1.02
USSO
|
Abstract
|
P00412 IBMPC 00 |
Computer Code System for the Estimation of Long-Term Performance of Cement-Based Materials. |
CEPXS/ONELD 1.0
|
Abstract
|
C00544 MNYCP 02 |
One-Dimensional Coupled Electron-Photon Multigroup Discrete Ordinates Code System. |
CERPI-CEREL |
Abstract
|
P00147 I0360 00 |
Code Systems for Automatic Analysis of Gamma-Ray Spectra Obtained with Ge(Li) Detectors. |
CGS 11.4 |
Abstract
|
P00243 MFMWS 03 |
Common Graphics System, Version 11.4. |
CHAINS-PC
|
Abstract
|
C00604 IBMPC 00 |
Code System to Compute Atom Density of Members of a Single Decay Chain. |
CHAINT-MC
|
Abstract
|
C00584 CYXMP 00 |
A Two-Dimensional Model for the Analysis of Contaminant Transport in a Fractured Porous Medium. |
CHARGE II
|
Abstract
|
C00070 C6500 00 |
Space Radiation Shielding Code - Proton and Electron Penetration of Multilayered Slabs and Spheres. |
CHARGE II
|
Abstract
|
C00070 I3675 00 |
Space Radiation Shielding Code - Proton and Electron Penetration of Multilayered Slabs and Spheres. |
CHARGE-PC
|
Abstract
|
C00070 IBMPC 00 |
Space Radiation Shielding Code - Proton and Electron Penetration of Multilayered Slabs and Spheres. |
CHENDF 7.02 |
Abstract
|
P00333 MNYCP 05 |
Codes for Handling ENDF/B-V and ENDF/B-VI Data. |
CHNSED
|
Abstract
|
C00671 I0360 00 |
Code System to Model Sediment & Containment Transport. |
CITATION-LDI 2
|
Abstract
|
C00643 PC386 02 |
Nuclear Reactor Core Analysis Code System. |
CLAW-IV |
Abstract
|
D00036 I0360 02 |
Coupled 30 Neutron, 12 Gamma-Ray Group Cross Sections Based on ENDF/B-IV for Radiation Transport Calculations. |
CLAW-IV |
Abstract
|
D00036 I3033 03 |
Coupled 30 Neutron, 12 Gamma-Ray Group Cross Sections Based on ENDF/B-IV for Radiation Transport Calculations. |
CLEAR |
Abstract
|
D00042 I3691 00 |
126 Neutron, 36 Gamma-Ray Cross Sections in AMPX and CCCC Interface Formats for LMFBR Neutronics Calculations. |
CLOUD-M
|
Abstract
|
C00032 I3565 00 |
Gamma-Ray Dose Rate from a Radioactive Cloud-Kernel Integration Code. |
CNCSN
|
Abstract
|
C00726 PC586 00 |
One, Two- and Three-Dimensional Coupled Neutral and Charged Particle SN Code System. |
COAG-II |
Abstract
|
P00070 I0360 00 |
Calculation of the Westcott Epithermal Index and the Westcott 2200 m/s Neutron Flux. |
COBB |
Abstract
|
D00016 I3675 01 |
123-Group Neutron Cross Section Data Generated from ENDF/B-II Data for Use in the XSDRN Discrete Ordinates Spectral Averaging Code. |
COBRA-EN |
Abstract
|
P00507 IBMPC 00 |
Code System for Thermal-Hydraulic Transient Analysis of Light Water Reactor Fuel Assemblies and Cores. |
COBRA-SFS CYCLE 3 |
Abstract
|
P00472 MNYCP 00 |
Code System for Thermal Hydraulic Analysis of Spent Fuel Casks. |
COBRA4I |
Abstract
|
P00419 MNYCP 00 |
Code Sytem to Calculate Rod-Bundle and Core Thermal-Hydraulics. |
CODAC (2) |
Abstract
|
P00073 I0360 00 |
For TIMOC 72, Monte Carlo Three-Dimensional Neutron Transport Code's Data Generator. |
COG 10
|
Abstract
|
C00724 MNYCP 00 |
Multiparticle Monte Carlo Code System for Shielding and Criticality Use. |
COGAP |
Abstract
|
P00375 MNYCP 01 |
Nuclear Power Plant Containment Hydrogen Control System Evaluation Code. |
COHORT-II
|
Abstract
|
C00198 I7094 00 |
General Purpose Monte Carlo Radiation Transport Code System. |
COLLIMATOR
|
Abstract
|
C00136 I7090 00 |
Monte Carlo Calculation of the Spectrum of Gamma Radiation from a Collimated Co-60 Source. |
COLUMN2
|
Abstract
|
C00534 ALLMF 00 |
Calculation of Effects of Physicochemical Processes on Migration. |
COMAND |
Abstract
|
P00091 I0360 00 |
A Multigroup ANISN Cross Section Data Library Collapsing Code System. |
COMBINE-PC |
Abstract
|
P00286 IBMPC 00 |
Code System to Compute Neutron Spectra and ENDF/B Version 5 Based Multigroup Neutron Constants. |
COMIDA |
Abstract
|
P00343 MNYCP 00 |
Radionuclide Food Chain Model for Acute Fallout Deposition. |
COMMIX-1B
USSO
|
Abstract
|
P00393 DVX11 00 |
3-D Single-Phase Thermal Hydraulics |
COMMIX-1B
USSO
|
Abstract
|
P00393 I3033 00 |
3-D Single-Phase Thermal Hydraulics |
COMMIX-1C
USSO
|
Abstract
|
P00393 MNYCP 00 |
3-D Single-Phase Thermal Hydraulics |
COMNUC3B |
Abstract
|
P00302 CYXMP 00 |
A Compound Nucleus Analysis Program. |
COMPAR |
Abstract
|
P00240 C0170 00 |
Compares Multigroup Cross Sections Generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS. |
COMPARE-MOD1A |
Abstract
|
P00410 C7600 00 |
Transient Flow W/Sinks & Doors |
COMPARE-MOD1A |
Abstract
|
P00410 I3033 00 |
Code System to Calculate Transient Flow With Heat Sinks & Doors. |
COMPASS 1.0.0 |
Abstract
|
P00520 PC586 00 |
Computerization of MARSSIM for Planning and Assessing Site Surveys. |
COMPBRN3 |
Abstract
|
P00389 PC386 00 |
Code System for Modeling Compartment Fires. |
COMPLOT |
Abstract
|
P00259 IBMMF 00 |
Convert EXFOR Format Data to Computation Format and Plot Comparisons of EXFOR and ENDF/B Evaluated Data (Version 86-1). |
COMPRASH
|
Abstract
|
C00072 I3675 00 |
Spinney Removal-Diffusion Shielding Code. |
COMRADEX4
|
Abstract
|
C00332 I0360 00 |
Evaluator of Potential Radiological Doses in the Near (< 10 km) Environment of Radioactive Release. |
CONDOS-II
|
Abstract
|
C00416 I0360 00 |
Code for Estimating Radiation Doses from Radionuclide-Containing Consumer Products. |
CONFOLD |
Abstract
|
P00053 C6600 00 |
Least-Structure Unfolding Code System for Measured Neutron and Gamma-Ray Spectra. |
CONFOLD |
Abstract
|
P00053 I0360 00 |
Least-Structure Unfolding Code System for Measured Neutron and Gamma-Ray Spectra. |
CONSTRIP V
|
Abstract
|
C00139 I3675 00 |
Vertical Barrier-Finite Source Plane Gamma-Ray Penetration Code System. |
CONTEMPT-LT28B
USSO
|
Abstract
|
P00387 C7600 00 |
Code System to Predict Containment Pressure-Temperature Response To a Loss-Of-Coolant Accident. |
CONTEMPT4 |
Abstract
|
P00397 MNYCP 00 |
Code System for PWR & BWR Multicompartment Containment Analysis, Versions MOD5 & MOD6. |
CONVERT |
Abstract
|
P00036 C6600 00 |
An IBM-to-CDC Program Conversion Code. |
COOL-C |
Abstract
|
P00017 I0360 00 |
Spectra Unfolding Codes. |
CORTES |
Abstract
|
P00404 I0360 00 |
Code System for Thermal & Mechanical Analysis of Tees. |
COVERV |
Abstract
|
D00077 I0360 01 |
Compilation of Multigroup Cross-section Covariance Matrices in COVERX Format for Several Important Materials (Generated from ENDF/B-V Data using PSR-093/PUFF2). |
COVERX |
Abstract
|
D00044 I0360 02 |
Compilation of Multigroup Cross-Section Covariance Matrices in COVERX Format for Several Important Materials. |
COVFILS |
Abstract
|
D00091 I0360 00 |
A 30-Group Covariance Library Based on ENDF/B-V. |
COVFILS-2 |
Abstract
|
D00137 ALLCP 00 |
Neutron Data and Covariances for Sensitivity and Uncertainty Analysis. |
CRAC2
|
Abstract
|
C00419 C0000 00 |
Code System for Calculating Reactor Accident Consequences. |
CRAC2
|
Abstract
|
C00419 I3033 00 |
Code System for Calculating Reactor Accident Consequences. |
CRECTJ5 |
Abstract
|
P00250 D0780 00 |
A Computer Program for Compilation of Evaluated Nuclear Data in ENDF/B Format. |
CRESO |
Abstract
|
P00184 I3081 00 |
Resonance Data-Handling Code System. |
CRRIS
|
Abstract
|
C00518 I3033 00 |
Computerized Radiological Risk Investigation System for Assessing Doses and Health Risks from Atmospheric Releases of Radionuclides. |
CRRIS
|
Abstract
|
C00518 PC586 00 |
Computerized Radiological Risk Investigation System for Assessing Doses and Health Risks from Atmospheric Releases of Radionuclides. |
CRRS |
Abstract
|
P00376 DALPU 01 |
Conference Room Reservation System. |
CRYSTAL BALL
|
Abstract
|
C00233 C6600 00 |
Code System for Determining Neutron Spectra from Activation Measurements. |
CRYSTAL BALL
|
Abstract
|
C00233 I0360 00 |
Code System for Determining Neutron Spectra from Activation Measurements. |
CTR DATA |
Abstract
|
D00028 I3675 01 |
73-Group P3 Coupled Neutron and Gamma-Ray Cross Sections for Fusion Reactor Calculations. |
CUPED |
Abstract
|
P00032 I3675 00 |
Scintillation Spectrometer Polyenergetic Gamma Photon Experimental Distributions Unfolding Code. |
CYGAS
|
Abstract
|
C00317 I3033 00 |
A Gamma-Ray Attenuation Code System for Large Gamma-Ray Sources Shielded by Coaxial Cylinders. |
CYGNUS-C SPHERE
|
Abstract
|
C00232 I0360 00 |
Monte Carlo Neutron Transport Code System in Spherical Geometry. |
CYLDOS
|
Abstract
|
C00389 I0360 00 |
A Cylindrical Geometry Gamma-Ray Flux Attenuation Code System. |
D2O |
Abstract
|
P00398 PC486 00 |
Code System for Computing Thermodynamic and Transport Properties of D2O. |
DABL69 |
Abstract
|
D00130 I0360 01 |
Defense Nuclear Applications Broad-Group Library based on ENDF/B-V in ANISN Format. |
DACRIN
|
Abstract
|
C00273 U1100 00 |
Airborne Radionuclide Organ Dose Calculational System. |
DANCOFF-MC |
Abstract
|
P00509 MNYCP 00 |
Code System for Monte Carlo Calculation of Dancoff Factors in Irregular Geometries. |
DANCOFF3 |
Abstract
|
P00279 D8810 00 |
Calculates Dancoff Correction. |
DANTE |
Abstract
|
P00185 I0370 00 |
Unfolding Code System for Energy Spectra Evaluation for Dosimetry Purposes. |
DANTSYS 3.0
|
Abstract
|
C00547 MFMWS 01 |
One-, Two-, and Three-Dimensional, Multigroup, Discrete-Ordinates Transport Code System. See new release CCC-707/PARTISN. |
DASH-FP
|
Abstract
|
C00366 C0000 00 |
A One-Dimensional Analytic-Numerical Solution to the Problem of Multicomponent Time-Dependent Diffusion of Fission Products. |
DASQHE |
Abstract
|
P00278 D8810 00 |
Calculates Dancoff Corrections Factors. |
DATINIT |
Abstract
|
P00258 DGMV1 00 |
Interactive Program To Access Photon Interaction Data. |
DAVE
|
Abstract
|
C00166 I3675 00 |
Monte Carlo Gamma-Ray Transport Code System in One-Dimensional Spherical Geometry. |
DCHAIN 1.3
|
Abstract
|
C00640 MNYCP 01 |
Code System for Radioactive Decay and Reaction Chain Calculations. |
DCHAIN-SP2001
|
Abstract
|
C00712 MNYWS 01 |
Code System for Analyzing Decay and Build-up Characteristics of Spallation Products. |
DCHAIN2
|
Abstract
|
C00370 PC486 00 |
A Code System for Calculation of Transmutation of Nuclides. |
DCTDOS
|
Abstract
|
C00520 IBMPC 00 |
Neutron and Gamma-Ray Penetration in Composite Duct Systems. |
DDXCODES
|
Abstract
|
C00583 FM380 00 |
One-, Two- and Three-Dimensional Transport Codes Using Multigroup Double-Differential Form Cross Sections. |
DDXLIB |
Abstract
|
D00123 FM380 01 |
125-Neutron Group Double Differential Cross Section Library. |
DECAYREM |
Abstract
|
D00030 I0360 02 |
Radioactive Decay Spectra in EXREM Format. |
DECDC 1.0 |
Abstract
|
D00213 MNYCP 00 |
Nucear Decay Data Files for Radiation Dosimetry Calculations. |
DEIS
|
Abstract
|
C00455 C6600 00 |
Draft Environmental Impact Statement on Licensing Requirements for Land Disposal of Radioactive Waste. |
DELFIC-TES
|
Abstract
|
C00257 I3033 00 |
Multigroup Kernel Integration Code System for Calculating Gamma-Ray Exposure from a Radioactive Airborne Cloud. |
DELFIC-TES
|
Abstract
|
C00257 U1108 00 |
Multigroup Kernel Integration Code System for Calculating Gamma-Ray Exposure from a Radioactive Airborne Cloud. |
DEMON & DEMON R
|
Abstract
|
C00181 I3675 00 |
Demonstration Monte Carlo Code System in Slab Geometry. |
DENIS |
Abstract
|
P00082 I0360 00 |
Monte Carlo Simulation of the Capture and Detection of Neutrons with Large Liquid Scintillators. |
DEPLETOR |
Abstract
|
P00523 MNYCP 00 |
Code System to Provide Depletion Capability to the U.S. NRC PARCS Code |
DEPOSITION
FEDC
|
Abstract
|
P00420 IBMPC 00 |
Code System to Calculate Particle Penetration Through Aerosol Transport Lines. |
DETAN 95 |
Abstract
|
P00361 MNYCP 00 |
Code System to Calculate Spectrum-Averaged Cross Sections and Detector Responses in Neutron Spectra. |
DIAMANT2
|
Abstract
|
C00414 PC386 00 |
Multigroup Two-Dimensional Discrete Ordinates Transport Code System for Triangular Geometry, Release 2.0. |
DIF3D8-VARIANT8
|
Abstract
|
C00649 MFMWS 01 |
Code System Using Variational Nodal Methods and Finite Difference Methods to Solve Neutron Diffusion and Transport Problems. |
DIFBAS |
Abstract
|
P00334 MNYCP 00 |
A Bayesian Approach to Unfolding a Neutron Spectrum from a Spectrum of Recoiled Protons. |
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. |
DIMEN |
Abstract
|
P00341 IBMPC 00 |
Code System for Isotope Identification by Gamma-Ray Analysis. |
DINT |
Abstract
|
P00049 C6600 00 |
Multigroup Coherent-Incoherent Cross Section Data Generator for Photon Transport Calculations. |
DINT |
Abstract
|
P00049 I0360 00 |
Multigroup Coherent-Incoherent Cross Section Data Generator for Photon Transport Calculations. |
DINT-YAEC
|
Abstract
|
C00306 ALLMF 00 |
Evaluator of I1 and I2 Integrals as Used in Long-Term External Gamma-Ray Doses from Routine Atmospheric Releases. |
DIPHO
|
Abstract
|
C00140 I3675 00 |
Monte Carlo Gamma-Ray Code System-Infinite Medium, Mono-energetic and Isotropic Point Source. |
DISDOS
|
Abstract
|
C00170 I0360 00 |
Calculation of Dose Distribution in Human Phantoms Irradiated by External Photon Sources. |
DISKTRAN
|
Abstract
|
C00533 CYXMP 00 |
Dose Calculations at Detectors from the End of a Cylinder Using DOT IV Scalar Flux Data. |
DISKTRAN
|
Abstract
|
C00533 I3033 00 |
Dose Calculations at Detectors from the End of a Cylinder Using DOT IV Scalar Flux Data. |
DISPERS
|
Abstract
|
C00454 MNYCP 00 |
Mathematical Models for Dispersion of Radionuclides |
DKR
|
Abstract
|
C00323 CY000 00 |
A Radioactivity and Dose Rate Calculation Code System for Fusion Reactors. |
DLS
|
Abstract
|
C00264 C6600 00 |
Two-Dimensional Shielding Calculational System with Diffusion Theory and Line-of-Sight Method. |
DOMINO |
Abstract
|
P00064 I0360 00 |
A General Purpose Code System for Coupling Discrete Ordinates and Monte Carlo Radiation Transport Calculations. |
DOMINO-II |
Abstract
|
P00162 I3033 00 |
General Purpose Code System for Coupling DOT-IV Discrete Ordinates and Monte Carlo Radiation Transport Calculations. |
DOMUS |
Abstract
|
P00301 IPCXT 00 |
A Program for Decomposing A Two-Dimensional Spectrum. |
DOORS 3.2A
|
Abstract
|
C00650 MFMWS 04 |
One, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System. |
DOPEX
|
Abstract
|
C00177 I3675 00 |
Laminated Shield Weight Optimization Code System-Steepest Descent Calculational Model. |
DOPEX
|
Abstract
|
C00177 U1108 00 |
Laminated Shield Weight Optimization Code System-Steepest Descent Calculational Model. |
DOPEX-1D2C
|
Abstract
|
C00214 I0360 00 |
A One-Dimensional, Two-Constraint Radiation Shield Optimization Code System. |
DOQDP |
Abstract
|
P00110 I0360 00 |
Discrete Ordinates Quadrature Generator. |
DORGLIB |
Abstract
|
P00181 I0360 00 |
An Interactive Program for Displaying Nuclide Decay and Generation Data Based on ORIGEN Data Library. |
DORIAN |
Abstract
|
P00425 IBMPC 00 |
Code System to Implement Bayes Method for Plant Aging Risk Analysis. |
DOSCOV |
Abstract
|
D00090 I0360 00 |
24-Group Covariance Data. |
DOSDAM77-81 |
Abstract
|
D00081 C6400 00 |
620 Group, SAND-II Formatted, Neutron Cross Sections Based on ENDF/B-IV and Other Sources for Spectral, Integral, and Damage Analyses. |
DOSDAM81-82 |
Abstract
|
D00097 C0000 00 |
Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses. |
DOSDAM84 |
Abstract
|
D00131 IBMMF 00 |
Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses. |
DOSDAT II-81 |
Abstract
|
D00079 I0370 00 |
Dose-Rate Conversion Factors for External Exposure to Photons and Electrons. |
DOSDAT-DOE |
Abstract
|
D00144 ALLMF 00 |
Dose-Rate Conversion Factors for External Exposure to Photons and Electrons. |
DOSDAT-DOE |
Abstract
|
D00144 IBMPC 01 |
Dose-Rate Conversion Factors for External Exposure to Photons and Electrons. |
DOSE 1
|
Abstract
|
C00165 I3675 00 |
Gamma-Radiation Dosimetry for Arbitrary Source and Target Geometry. |
DOSE-SGTR
|
Abstract
|
C00624 IBMPC 00 |
Code System to Calculate the Integrated Iodine Release to the Environment During a Steam Generator Tube Rupture in a PWR. |
DOSFACTER II
|
Abstract
|
C00400 D0750 00 |
Calculation of Dose-Rate Conversion Factors for Exposure to Photons and Electrons. |
DOSFACTER II
|
Abstract
|
C00400 I0360 00 |
Calculation of Dose-Rate Conversion Factors for Exposure to Photons and Electrons. |
DOSFACTER-DOE
|
Abstract
|
C00536 I3033 00 |
Calculation of Dose-Rate Conversion Factors for Exposure to Photons and Electrons. |
DPCT
|
Abstract
|
C00580 CYXMP 00 |
A Deterministic-Probabilistic Model For Contaminant Transport. |
DPL-400 GEDT1 |
Abstract
|
D00031 I0360 08 |
Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
DPL-401 NEDT |
Abstract
|
D00031 I0360 09 |
Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
DPL-402A/GPDT1 |
Abstract
|
D00031 I0360 10 |
Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
DPL-402B/GPDT1 |
Abstract
|
D00031 I0360 11 |
Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
DRAGON3.05D
|
Abstract
|
C00647 MNYWS 03 |
Lattice Cell Code System. |
DRALIST |
Abstract
|
D00080 ALLCP 00 |
Radioactive Decay Data for Application to Radiation Dosimetry and Radiological Assessments. |
DSNQUAD |
Abstract
|
P00251 IPCXT 00 |
Calculates Angular Quadrature Weights and Cosines. |
DTF-69
|
Abstract
|
C00130 C6600 00 |
One-Dimensional Multigroup Photon Transport Discrete Ordinates Code System. |
DTF-INDIA
|
Abstract
|
C00458 I0370 00 |
Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering. |
DTF-IV
|
Abstract
|
C00042 C6600 00 |
Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering. |
DTF-IV MODIFIED
|
Abstract
|
C00042 I0370 00 |
Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering. |
DTF-TRACA
|
Abstract
|
C00412 U1100 00 |
One-Dimensional Multigroup Neutron Transport Discrete Ordinates Code System. |
DTK
|
Abstract
|
C00223 I3675 00 |
One-Dimensional Multigroup Neutron Transport Code System. |
DUFOLD |
Abstract
|
P00042 I0360 00 |
Derivative Unfolding Code - Determination of Neutron Spectra from NE-213 Pulse Height Data. |
DUST
|
Abstract
|
C00453 I3033 00 |
Albedo Monte Carlo Simulation of Neutron Streaming Through Multilegged Ducts. |
DUST-BNL
|
Abstract
|
C00634 PC386 00 |
Disposal Unit Source Term by One-Dimensional, Transient, Finite-Difference, Subsurface Release and Transport of Contaminants. |
DWBA07/DWBB07 |
Abstract
|
P00338 MNYCP 01 |
Code System for Inelastic and Elastic Scattering with Nucleon-Nucleon Potential |
DWNWND
|
Abstract
|
C00383 DP010 00 |
Interactive Gaussian Plume Atmospheric Transport Model. |
DWUCK-CHUCK |
Abstract
|
P00546 MNYCP 00 |
Nuclear Model Code System for Distorted Wave Born Approximation and Coupled Channel Calculations. |
E-DEP-1
|
Abstract
|
C00275 D0VAX 00 |
Heavy Ion Energy Deposition Code System. |
E3LWR |
Abstract
|
D00098 C0000 00 |
45 Neutron, 16 Gamma-Ray and 15 Neutron, 5 Gamma-Ray Group LWR Cross Section Libraries Derived from EURLIB-III using the AGRUKO Optimized Collapsing Scheme. |
EASY-2005.1
|
Abstract
|
C00735 MNYCP 01 |
Inventory Code System for Neutron Activation Analysis. |
EASY-QAD 1.0
|
Abstract
|
C00744 PC586 00 |
Visualization Code System for Gamma and Neutron Shielding Calculations. |
ECIS-06 |
Abstract
|
P00227 MNYCP 01 |
Code System to Solve the Coupled Differential Equations Arising in Nuclear Model Calculations. |
ECPL82 |
Abstract
|
D00106 ALLCP 00 |
Evaluated Charged-Particle Data Library. |
EDISTR |
Abstract
|
P00191 I3033 00 |
Prepares a Nuclear Decay Data Base for Internal Radiation Dosimetry Calculations. |
EDITOR |
Abstract
|
P00035 I0360 00 |
Alters Mode, Copies, Merges, Punches, Edits, or Adds to ENDF/B-Formatted Data on Tapes or Cards. |
EDMULT 6.4
|
Abstract
|
C00430 MNYCP 02 |
Evaluates Electron Depth-Dose Distributions in Multilayer Slab Absorbers. |
EDNA
|
Abstract
|
C00104 I7090 00 |
Electron Dose and Number Analysis Code by Kernel Integration. |
EDO
|
Abstract
|
C00489 U1110 00 |
A Code System in Fortran V for the Evaluation of Dose During Normal Operation of a Nuclear Power Plant. |
EDSFI
USSO
|
Abstract
|
D00215 PC486 00 |
Electrical Distribution System Functional Inspection Data Base. |
EEDB |
Abstract
|
P00531 MNYCP 00 |
The Energy Economic Data Base. |
EFDOS
|
Abstract
|
C00411 I0360 00 |
Calculation of Effective Committed Dose Equivalents by Inhalation of Radioactive Materials Occurring in Routine Atmospheric Releases from Nuclear Fuel Cycle Facilities. |
EGAD
|
Abstract
|
C00206 I0360 00 |
Calculation of Dose from External Gamma-Ray Emitters. |
EGS4
|
Abstract
|
C00331 MNYCP 00 |
Monte Carlo Simulation of the Coupled Transport of Electrons and Photons. |
ELAN |
Abstract
|
P00141 ICL00 00 |
Neutron Cross-Section Self-Shielding Code System. |
ELAST2 |
Abstract
|
D00208 MNYCP 00 |
Database of Cross Sections for the Elastic Scattering of Electrons and Positrons by Atoms. |
ELBA
|
Abstract
|
C00119 I0360 00 |
Electron and Bremsstrahlung Dose Rate Code. |
ELECSPEC |
Abstract
|
D00100 DP010 00 |
Electron Spectra from Decay of Fission Products. |
ELF
|
Abstract
|
C00167 I0360 00 |
Monte Carlo Neutron Transport Code System for Cylinders and Spheres. |
ELGATL
|
Abstract
|
C00295 C6600 00 |
Calculation of Energy Spectra from Coupled Electron-Photon Slowing Down. |
ELIESE-3 |
Abstract
|
P00003 I0370 00 |
Analyses of Elastic and Inelastic Scattering Cross Sections. |
ELPHO
|
Abstract
|
C00301 I0360 00 |
Three-Dimensional Monte Carlo Electromagnetic Transport Code System. |
ELTRAN
|
Abstract
|
C00155 C3600 00 |
One-Dimensional Monte Carlo Electron Transport Code System. |
EMERALD
|
Abstract
|
C00211 I0360 00 |
Calculation of Activity Releases and Potential Doses from a Pressurized Water Reactor Plant. |
EMERALD-NORMAL
|
Abstract
|
C00250 I0370 00 |
Calculation of Activity Releases and Potential Doses from the Normal Operation of a Pressurized Water Reactor Plant. |
EMPIRE |
Abstract
|
P00292 IPCAT 00 |
A Pre-equilibrium Compound Nuclear Model Code For Personal Computers. |
EMPIRE-II |
Abstract
|
P00497 PC586 00 |
Statistical Model Code System for Nuclear Reaction Calculations. |
ENBAL2 |
Abstract
|
P00160 I0370 00 |
A Program to Generate Multigroup Neutron Kerma Factors. |
ENDL82 |
Abstract
|
D00103 ALLCP 00 |
Neutron Library in Transmittal Format. |
ENDLIB-97 |
Abstract
|
D00179 MNYCP 01 |
LLNL Libraries of Atomic Data, Electron Data, and Photon Data in Evaluated Nuclear Data Library (ENDL) Type Format. |
ENEDEP
|
Abstract
|
C00227 GE400 00 |
Energy Deposition Code System for GE 265 Time-Sharing System. |
ENLOSS |
Abstract
|
P00047 C6600 00 |
Calculation of Energy Loss of Charged Particles. |
ENSL82-CDRL82 |
Abstract
|
D00110 ALLCP 00 |
Evaluated Nuclear Structure Libraries. |
ENTOSAN |
Abstract
|
P00188 C0175 00 |
Code System for Calculating Fine-Group Dosimetry Cross Section Values from ENDF/B Data. |
ENTOSAN |
Abstract
|
P00188 D8810 00 |
Code System for Calculating Fine-Group Dosimetry Cross Section Values from ENDF/B Data. |
ENTREE 1.4.0 |
Abstract
|
P00519 MNYWS 00 |
BWR Core Simulation System for Space and Time Dependent Coupled Phenomena. |
EPIPE
USSO
|
Abstract
|
P00485 CY000 00 |
Code System for Static and Dynamic Piping System Analysis. |
EPR |
Abstract
|
D00037 I3691 05 |
Coupled 100-Group Neutron 21-Group Gamma-ray Cross Sections for EPR Neutronics. |
EPR MASTER |
Abstract
|
D00052 I3691 00 |
100 Neutron Group Cross Sections in AMPX Master Library Format. |
EPRI-CINDER
|
Abstract
|
C00309 C6600 00 |
General Point-Depletion and Fission Product Code System and Four-Group Fission Product Neutron Absorption Chain Data Library Generated from ENDF/B-IV for Thermal Reactors. |
EQUIVA-1.1 |
Abstract
|
P00323 IMFPC 00 |
Generation of Few-Group Equivalent Diffusion Theory Parameters for PWR Reflector Regions. |
EQUIVA-2 |
Abstract
|
P00324 IMFPC 00 |
Generation of Environment-Insensitive Equivalent Diffusion Theory Parameters for PWR Reflector Regions. |
ERANOS 2.0
OECD
|
Abstract
|
C00745 MNYWS 00 |
Modular Code and Data System for Fast Reactor Neutronics Analyses |
ERIC-2 |
Abstract
|
P00119 I0360 00 |
Calculator of Resonance Integral and Effective Capture and Fission Cross Sections for Fissile and Non-Fissile Nuclides - Thermal or Fast Reactors. |
ERINNI |
Abstract
|
P00219 I0360 00 |
Optical Model Calculation of Multiple Cascading Particle Emissions. |
ERPEX
|
Abstract
|
C00305 C0073 00 |
Monte Carlo Distributions of Energetic Proton Ranges in Silicon. |
ERRORJ |
Abstract
|
P00526 MNYCP 01 |
Multigroup Covariance Matrices Generation from ENDF/B-6 Format. |
ESDORA
|
Abstract
|
C00183 U1108 00 |
Fission Product Inventory and Gamma-Ray Dose Rate from a Radioactive Cloud System. |
ESG |
Abstract
|
D00065 I0360 00 |
56-Group Cross Section Library Based on VITAMIN-C Generated by Using SPHINX and XSDRNPM to Collapse 171 Groups. |
ESP
|
Abstract
|
C00193 I0360 00 |
General Purpose Monte Carlo Neutron Transport Code System. |
ESTIMA |
Abstract
|
P00201 I3033 00 |
A Code System for Calculating Average Parameters from Sets of Resolved Resonance Parameters. |
ETHEL |
Abstract
|
P00217 I0360 00 |
Code System for Generating Cross Sections for PSR-128/THERMOS. |
ETRAN
|
Abstract
|
C00107 I0360 00 |
Monte Carlo Code System for Electron and Photon Through Extended Media. |
EURCYL |
Abstract
|
P00076 I0370 00 |
Finite Element Three-Dimensional Mesh Generator for Cylinder - Cylinder Intersections. |
EURLIB-III |
Abstract
|
D00035 I0360 01 |
100 Neutron, 20 Gamma-Ray Group Cross Section Library for Use in the European Shielding Benchmark Program. |
EVALPLOT |
Abstract
|
P00211 I3081 00 |
A Program to Plot Data in the Evaluated Nuclear Data File/Version B Format. |
EVAP |
Abstract
|
P00010 I0360 00 |
Calculation of Particle Evaporation from Excited Compound Nuclei. |
EVNTRE |
Abstract
|
P00465 D0VAX 00 |
Code System for Event Progression Analysis for PRA. |
EXIFON-GAMMA |
Abstract
|
P00305 IPCXT 00 |
A Model For Statistical Multistep Direct and Multistep Compound Reactions. |
EXPRESS
|
Abstract
|
C00622 MNYCP 00 |
Exact Preparedness Supporting System. |
EXTREME
|
Abstract
|
C00440 I3033 00 |
Two-Dimensional Discrete-Ordinates Code System with Exponential Expansion of Spatial Variables. |
EZVIDEO |
Abstract
|
P00237 IBMPC 00 |
Graphics Routines for the IBM PC. |
F5TAB |
Abstract
|
P00221 D0780 00 |
Code System for Converting Energy Distribution Cross Section Data to Tabulated Data. |
FAMREC |
Abstract
|
P00167 C7600 01 |
Fuel Assembly Mechanical Response Code System. |
FANAC |
Abstract
|
P00179 I3033 00 |
A Shape Analysis Code Package for Resonance Parameter Extraction from Neutron Capture Data for Light- and Medium-Weight Nuclei. |
FANAL |
Abstract
|
P00178 I3033 00 |
A Least-Squares Shape Analysis Code System. |
FANG |
Abstract
|
P00140 C0000 00 |
An Angular Folding Code System for Channel Theory Analysis. |
FANG |
Abstract
|
P00140 I0360 00 |
An Angular Folding Code System for Channel Theory Analysis. |
FANTOM
|
Abstract
|
C00375 BESM6 00 |
Monte Carlo Calculation of the Response of an External Detector to a Photon Source in the Lungs of a Heterogeneous Phantom. |
FASTER III
|
Abstract
|
C00168 U1108 00 |
Monte Carlo Neutron and Photon Transport Code System in Complex Geometries. |
FASTER-III
|
Abstract
|
C00168 I3675 00 |
Monte Carlo Neutron and Photon Transport Code System in Complex Geometries. |
FASTGRASS |
Abstract
|
P00479 MNYCP 00 |
Code System to Predict Fission Product Release in Ubase Fuels. |
FASTPLOT 1.0 |
Abstract
|
P00354 IBMPC 00 |
Interface to Microsoft FORTRAN Graphics. |
FATDUD |
Abstract
|
P00080 I0360 00 |
Foil Activation Data Unfolding Code System. |
FBSAM |
Abstract
|
P00103 I0360 00 |
User-Storage - Magnetic Disk Data Manipulator. |
FCXSEC |
Abstract
|
D00085 PC386 01 |
22 Neutron, 21 Gamma-Ray Group Cross Section Libraries in ANISN Format for Nuclear Fuel Cycle Shielding Calculations. |
FDKR
|
Abstract
|
C00541 I4381 00 |
Radioactivity and Dose Rate Calculation Code for Fission, Fusion and Hybrid Reactors. |
FDMXPC |
Abstract
|
P00322 IPCAT 00 |
Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format. |
FE3DGW
|
Abstract
|
C00531 D0780 00 |
Code System for Finite-Element, Three-Dimensional Ground-Water Flow Analysis. |
FEDGROUP-3 |
Abstract
|
P00123 I0360 00 |
Program System for Processing Evaluated Nuclear Data in ENDF/B, KEDAK or UKNDL Format to Constants to be Used in Reactor Physics Calculation. |
FEDGROUP-R |
Abstract
|
P00349 MNYCP 00 |
Multigroup Neutron Cross Section Processing System from Data in ENDF/B Format. |
FEDGROUPC86REV3 |
Abstract
|
P00194 MNYCP 01 |
Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation. |
FEM-2D
|
Abstract
|
C00260 C6600 00 |
Two-Dimensional Diffusion Theory Code System Based on the Method of Finite Elements. |
FEMAXI 6 VER.1 |
Abstract
|
P00536 IBMPC 00 |
Code System for Light Water Reactor Fuel Analysis. |
FEMB
|
Abstract
|
C00340 B6700 00 |
A Two-Dimensional Diffusion Theory Finite Element Program. |
FEMRZ
|
Abstract
|
C00342 F2307 00 |
A Finite-Element Method Two-Dimensional Multigroup Neutron Transport Code System, (r,z) Geometry. |
FEMWASTE/FEMWATER
|
Abstract
|
C00451 C7600 00 |
A Finite-Element Model of Waste and Water Transport through Porous Saturated-Unsaturated Media. |
FEMWASTE/FEMWATER
|
Abstract
|
C00451 PC386 00 |
A Finite-Element Model of Waste and Water Transport through Porous Saturated-Unsaturated Media. |
FENDL-2.0 |
Abstract
|
D00183 MNYCP 01 |
Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications. |
FENDL-2.1 |
Abstract
|
D00222 MNYCP 00 |
Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications. |
FEP 4.16 |
Abstract
|
P00440 IBMPC 00 |
Fault-tree, Event tree, & P&ID Editors. |
FERD-PC |
Abstract
|
P00273 IBMPC 00 |
Interactive Multichannel Neutron and Gamma-Ray Spectrum Matrix Unfolding Code System. |
FERDO/FERD |
Abstract
|
P00102 I3033 00 |
Multichannel Neutron and Gamma-Ray Spectrum Matrix Unfolding Code Systems. |
FERDOR |
Abstract
|
P00017 I7090 00 |
Spectra Unfolding Codes. |
FERDOR |
Abstract
|
P00017 U1108 00 |
Spectra Unfolding Codes. |
FERRET |
Abstract
|
P00145 U0000 00 |
Least-Squares Solution to Nuclear Data and Reactor Physics Problems. |
FESH
|
Abstract
|
C00676 CDCMF 00 |
X-Y Multigroup Neutron Transport Code System. |
FEWA-FEMA
|
Abstract
|
C00477 I3033 00 |
A Finite Element Model of Water and Other Material through Aquifers. |
FEWG1-81 |
Abstract
|
D00031 I0370 06 |
Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
FEWG1-85 |
Abstract
|
D00031 I0360 07 |
Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
FGR-DOSE |
Abstract
|
D00167 ALLCP 01 |
Dose Coefficients from Federal Guidance Reports 11 and 12. |
FGXRRS |
Abstract
|
D00132 C0000 00 |
Few Group Cross Section Library for Research Reactor Calculations. |
FIGERO |
Abstract
|
P00149 C0000 00 |
Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations. |
FINELM
|
Abstract
|
C00483 MFMWS 00 |
Multigroup Finite Element Diffusion Code System. |
FIPDIG
|
Abstract
|
C00251 I0360 00 |
One-Dimensional Time-Dependent Fission Product Diffusion Code System. |
FIPDOR |
Abstract
|
D00068 I3691 00 |
126 Neutron Group Fission Product Cross Sections. |
FIRAC |
Abstract
|
P00444 CY000 00 |
Nuclear Facilities Fire Accident Model |
FIREDATA |
Abstract
|
D00125 PC486 00 |
Nuclear Power Plant Fire Data Base for Personal Computers. |
FIS-PROD |
Abstract
|
D00152 ALLCP 00 |
Chinese Evaluated Fission Product Yield Library in ENDF/B-V Format. |
FISP-6
|
Abstract
|
C00538 I3090 00 |
An Enhanced Code for the Evaluation of Fission Product Inventories and Decay Heat. |
FISPIN
|
Abstract
|
C00413 ICL00 00 |
Nuclide Inventory Calculation System. |
FISSP & CLOUD
|
Abstract
|
C00163 MNYCP 01 |
Fission Product Inventory, Release, Transport and Dose Calculation. |
FITOCO |
Abstract
|
P00189 C0175 00 |
Converter of Fine-Group Flux Density and Cross Section Data to Coarse Group Values. |
FLEP |
Abstract
|
D00022 I3033 00 |
Coefficients for the Analytic Representation of Nonelastic Cross Sections and Particle-Emission Spectra from Various Nucleon-Nucleus Collisions in the Energy Range 25 to 400 MeV. |
FLODIS |
Abstract
|
P00417 I0360 00 |
Code System to Calculate Thermal Response of FSV HTGR Core. |
FLOWPLOT II |
Abstract
|
P00234 I3033 00 |
Fluid Dynamics and Heat Transfer Plotting Package. |
FLUKA-TRANKA
|
Abstract
|
C00207 C6600 00 |
Three-Dimensional High-Energy Extranuclear Hadron Cascade Monte Carlo System for Cylindrical Backstop Geometries. |
FLUNG |
Abstract
|
D00086 I3033 00 |
Coupled 35-Group Neutron and 21-Group Gamma Ray, P3 Cross Sections for Fusion Applications. |
FLUSH |
Abstract
|
P00043 C6600 00 |
Spectral Unfolding Code - Stepwise Regression of System Response Functions. |
FLYSPEC-SHORTS |
Abstract
|
P00196 C7600 00 |
Neutron Unfolding Code System for Reducing Proton-Recoil Pulse-Height Obtained with NE-213 Liquid Scintillator. |
FOCUS
|
Abstract
|
C00390 I3033 00 |
Adjoint Monte Carlo Neutron Transport Code System. |
FONTA
|
Abstract
|
C00423 S4044 00 |
Code System For Calculating Individual And Collective Doses From Reactor Accidents Using Pasquill's Plume Model. |
FOOD
|
Abstract
|
C00403 U1108 00 |
Calculation of Radiation Dose to Man from Radionuclides in the Environment. |
FORECAST V3.0 |
Abstract
|
P00384 IBMPC 00 |
Forecast Regulatory Effects Cost Analysis Program. |
FORIST |
Abstract
|
P00092 C0000 00 |
Neutron Spectrum Unfolding Code System - Iterative Smoothing Technique. |
FORIST |
Abstract
|
P00092 I0360 00 |
Neutron Spectrum Unfolding Code System - Iterative Smoothing Technique. |
FORSEN |
Abstract
|
P00170 I0360 00 |
A Multigroup Processing Code for Use with Sensitivity Profiles to Assess the Effect of Cross Section Changes. |
FORSIM VI |
Abstract
|
P00078 C6600 00 |
A Fortran-Oriented Simulation Package for the Automated Solution of Partial and Ordinary Differential Equation Systems. |
FORSS
|
Abstract
|
C00334 C0000 00 |
A Sensitivity and Uncertainty Analysis Code System. |
FORSS
|
Abstract
|
C00334 I0360 00 |
A Sensitivity and Uncertainty Analysis Code System. |
FOTELP-2K6
|
Abstract
|
C00581 MNYCP 03 |
Monte Carlo Simulation of Photons, Electrons and Positrons Transport. |
FOURACES |
Abstract
|
P00183 I0370 00 |
Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries. |
FPDL |
Abstract
|
D00066 I0360 00 |
Fission Product Yields, Gamma Ray and Beta Spectra in ENDF-III Format for 235U, 238U, 239Pu, 232Th, and 233U. |
FPGAM
|
Abstract
|
C00386 F2307 00 |
Calculation of Fission-Product Gamma-Ray Spectra. |
FPIC
|
Abstract
|
C00028 I3675 00 |
Fission Product Inventory Code. |
FPIP
|
Abstract
|
C00162 C6600 00 |
Fission Product Inventory Code System. |
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. |
FRANTIC3 |
Abstract
|
P00406 CDCMF 00 |
Time-Dependent Reliability Analysis. |
FRAPCON2
USSO
|
Abstract
|
P00517 MFMWS 00 |
Fuel Rod Thermal-Mechanical Behavior, Versions FRAPCON2, FRAPCON2/VIM4, & FRAPCON2/VIM5. |
FRAPT6/MOD1
USSO
|
Abstract
|
P00436 C0176 00 |
Code System for Transient Analysis of Fuel Rods. |
FRAPT6/V21
USSO
|
Abstract
|
P00436 C0176 01 |
Code System for Transient Analysis of Fuel Rods. |
FRCRL2
|
Abstract
|
C00231 C6400 00 |
Calculation of Fission-Product Release in Reactor Accident Analyses. |
FREEFORM |
Abstract
|
P00081 I0360 00 |
Free-Form Input Reading Routines. |
FSCATT
|
Abstract
|
C00186 I3033 00 |
Discrete Ordinates Gamma-Ray Transport Code System in Plane Geometry. |
FSCATT
|
Abstract
|
C00186 U1108 00 |
Discrete Ordinates Gamma-Ray Transport Code System in Plane Geometry. |
FSX96 |
Abstract
|
D00190 MNYWS 00 |
Collection of Continuous Energy Cross Section Libraries for MCNP Based on JENDL 3.2, JENDL, Fusion File and Dosimetry File. |
FSXLIB-J3 |
Abstract
|
D00165 ALLCP 00 |
MCNP continuous energy neutron cross section library based on JENDL-3. See DLC-190/FSX96 based on JENDL3.2. |
FSXLIB-J33 |
Abstract
|
D00223 MNYCP 01 |
Continuous Energy Neutron Cross Section Library for MCNP Based on JENDL 3.3. |
FTF |
Abstract
|
D00056 I0360 00 |
Multigroup Neutron and Gamma-Ray Dose Transmission Factors for Concrete Slabs. |
FUELSDATA |
Abstract
|
P00446 C7600 00 |
Code System to Model Verification Fuel Rod Data. |
FURNACE
|
Abstract
|
C00615 C0740 00 |
Code System for Neutronic Calculations in Three Dimension Toroidal Geometry. |
G3-6ED
|
Abstract
|
C00075 C6600 00 |
Kernel Integration Code System - Multigroup Gamma Ray Scattering. |
G3-6ED
|
Abstract
|
C00075 I3033 00 |
Kernel Integration Code System - Multigroup Gamma Ray Scattering. |
G33-GP
|
Abstract
|
C00494 IBMPC 01 |
Kernel Integration Code System - Multigroup Gamma-Ray Scattering Using the GP Buildup Factor. |
GABAS |
Abstract
|
P00175 U1108 00 |
A Code System for Generating Composite Time-Dependent Fission Produce Spectra. |
GADJET
|
Abstract
|
C00115 C6600 00 |
Monte Carlo Gamma-Ray Adjoint Energy Transport Code in Complex Three-Dimensional Geometry. |
GAINCALB |
Abstract
|
P00056 I0360 00 |
Determination of the Gain Used with Organic Scintillation Detect. |
GALAXY-6 |
Abstract
|
P00098 I0370 00 |
Neutron Multigroup Cross Section Processor. |
GALE BWR
|
Abstract
|
C00335 U1100 00 |
Boiling Water and Pressurized Water Reactors Gaseous and Liquid Effluents Radiological Assessment Code System. |
GALE PWR & BWR
|
Abstract
|
C00335 I3033 00 |
Boiling Water and Pressurized Water Reactors Gaseous and Liquid Effluents Radiological Assessment Code System. |
GALE86
|
Abstract
|
C00506 C0000 01 |
Calculation of Routine Radioactive Releases in Gaseous and Liquid Effluents from Boiling Water and Pressurized Water Reactors. |
GAMAN |
Abstract
|
P00083 DP010 00 |
Qualitative and Quantitative Evaluation of Ge(Li) Gamma-Ray Spectra. |
GAMANAL |
Abstract
|
P00506 D0VAX 00 |
Code System for Computerized Quantitative Analysis By Gamma-Ray Spectrometry. |
GAMDAT-78 |
Abstract
|
D00083 I0370 00 |
Library of Gamma-Ray Decay Data for 2055 Radionuclides. |
GAMIDENT |
Abstract
|
P00154 C0000 00 |
A Program to Aid in the Identification of Unknown Materials by Gamma-ray Spectroscopy. |
GAMLEG-75 |
Abstract
|
P00086 C7600 00 |
Multigroup Cross Section Generator for Photon Transport Calculations. |
GAMLEG-JR |
Abstract
|
P00116 F2307 00 |
Multigroup Cross-Section Generator for Photon Transport Calculations. |
GAMLEG-JR |
Abstract
|
P00116 I3033 00 |
Multigroup Cross-Section Generator for Photon Transport Calculations. |
GAMLIB |
Abstract
|
D00006 I0360 00 |
99-Group Neutron Cross Sections for Use in the GAM Portion of the GGC Multigroup Cross Section Code. |
GAMMA |
Abstract
|
P00095 I0360 00 |
Monte Carlo Code System for Calculating Efficiencies and Response Functions of NaI(Tl) Crystals for Gamma Rays from Thick Disk Sources. |
GAMMOM
|
Abstract
|
C00135 ALLMF 00 |
Gamma-Ray Moments Method Codes--GRMM and SPENCER. |
GAMMOM-I
|
Abstract
|
C00226 I0360 00 |
Gamma-Ray Moments Method Code System. |
GAMMON |
Abstract
|
D00071 ALLCP 00 |
Activation Library for Fusion Reaction Application and Other Design Studies. |
GAMTAB |
Abstract
|
D00032 I0360 00 |
Radioactive-Decay Gamma-Rays Ordered by Energy and Nuclide. |
GAMTOT78 |
Abstract
|
D00109 CY00I 00 |
Compilation of Radioactive Decay and Capture Gamma Rays. |
GAMX1 |
Abstract
|
P00209 I0370 00 |
A Computer Code System for Evaluating Spectra Peak Areas. |
GANAPOL-ABNTT
|
Abstract
|
C00753 MNYCP 00 |
Analytical Benchmarks; Case Studies in Neutron Transport Theory. |
GAPCON-THERMAL |
Abstract
|
P00499 C7600 00 |
Code System to Calculate Fuel Steady State & Transient Behavior. |
GARG |
Abstract
|
D00073 C0000 00 |
27-Group Neutron Cross Sections in Discrete Ordinates Format Generated with FIGERO (PSR-149) from ENDF-B Data. |
GARLIB |
Abstract
|
D00013 I7090 00 |
Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations. |
GARLIB |
Abstract
|
D00013 I3565 01 |
Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations. |
GAROL |
Abstract
|
P00033 I7090 00 |
Calculation of Resonance Neutron Absorption in Two-Region Problems. |
GASPAR
|
Abstract
|
C00463 I3033 01 |
Calculates Radiation Exposure to Man from Routine Air Releases of Nuclear Reactor Effluents. |
GASPAR II
|
Abstract
|
C00463 D0780 00 |
Calculates Radiation Exposure to Man from Routine Air Releases of Nuclear Reactor Effluents. |
GASS
|
Abstract
|
C00080 I7090 00 |
Monte Carlo Calculation of Self Shielding by Encapsulated Gamma-Ray Sources. |
GAUSS V |
Abstract
|
P00045 I0360 00 |
A Code system for Analysis of Gamma-Ray Spectra from Ge(Li) Spectrometers. |
GAUSS VII |
Abstract
|
P00045 C0000 00 |
A Code system for Analysis of Gamma-Ray Spectra from Ge(Li) Spectrometers. |
GBANISN
|
Abstract
|
C00628 IRISC 00 |
One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering with the GroupBand Option. |
GCI |
Abstract
|
P00421 IBMPC 00 |
Generic Communications Index |
GEAF-1 |
Abstract
|
D00158 D8810 00 |
100 Group Cross Sections for Neutron Activation. |
GECINX |
Abstract
|
P00193 H6000 00 |
A Code System for Collapsing Multigroup Cross Sections in CCCC Format. |
GELI2/SPAN2 |
Abstract
|
P00094 I0360 00 |
Calculation of Nuclide Abundaces from Multichannel Gamma-ray Spectra. |
GENII 2.06
|
Abstract
|
C00737 PC586 00 |
Environmental Radiation Dosimetry Software System. |
GENII-LIN 2.1
|
Abstract
|
C00728 PC586 01 |
GENII-LIN Multipurpose Health Physics Code System with a New Object-Oriented Interface, Release 2.0. |
GENP-2
|
Abstract
|
C00575 ALLMF 00 |
Generalized Perturbation Theory Code System. |
GENRD |
Abstract
|
P00040 C6600 00 |
Free Format Card Input Processor. |
GENRD |
Abstract
|
P00040 I0360 00 |
Free Format Card Input Processor. |
GERES |
Abstract
|
P00241 I0370 00 |
A Code to Produce Cross-Section Libraries for ANISN Based on Heterogeneous Fast Reactor Cell Calculations Using MC2II Data. |
GES_MC
|
Abstract
|
C00742 PC586 00 |
Gamma-electron Efficiency Simulator, Version 3.1 |
GETOUT
|
Abstract
|
C00461 C0176 00 |
A Computer Code System for Predicting One-Dimensional Radionuclide Decay Chain Transport through Geologic Media. |
GFX-GAMIX
|
Abstract
|
C00397 I3033 00 |
A Spherical Harmonics Code System for Evaluation of Terrestrial Gamma-Radiation Fields. |
GGC-3 |
Abstract
|
P00012 I3565 00 |
Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
GGC-3 & GGC-4 |
Abstract
|
P00012 I3675 00 |
Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
GGC-4 |
Abstract
|
P00012 U1108 00 |
Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
GGG-GP
|
Abstract
|
C00564 IBMPC 00 |
Kernel Integration Code System - Multigroup Gamma-Ray Scattering Using the GP Buildup Factor. |
GGTC-ENEL |
Abstract
|
P00128 I0360 00 |
Code System for Producing Few-Group Neutron Cross Sections from Multigroup Data Libraries. |
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. |
GIFT |
Abstract
|
P00124 C0076 00 |
A Combinatorial Geometry Code System with Model Testing Routines. |
GIFT |
Abstract
|
P00124 D0VAX 00 |
A Combinatorial Geometry Code System with Model Testing Routines. |
GIFT |
Abstract
|
P00124 U0000 00 |
A Combinatorial Geometry Code System with Model Testing Routines. |
GIP |
Abstract
|
P00229 IBMPC 00 |
Group-Organized Cross-Section Input Program. |
GIRAFFE |
Abstract
|
P00304 I3033 00 |
General Isotope Release Analysis For Failed Elements. |
GLUCS |
Abstract
|
P00192 D0VAX 00 |
A Generalized Least-Squares Code System for Updating Cross Section Evaluations with Correlated Data Sets. |
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, Version gn9cp8. |
GNOMER
|
Abstract
|
C00625 MNYCP 01 |
Multigroup 3-Dimensional Neutron Diffusion Nodal Code System with Thermohydraulic Feedbacks. |
GOFRR |
Abstract
|
P00127 I0360 00 |
Generator of Graphical Output of DOT and ANISN Fluxes and Reaction Rates. |
GRACE-II
|
Abstract
|
C00026 I3675 00 |
Gamma Ray Kernel Integration Dose Rate and Heating Code-Cylinders and Spheres. |
GRASS-SST |
Abstract
|
P00489 MNYCP 00 |
Code System to Predict Fission-Gas Release & Fuel Swelling. |
GREAT-GRASS
|
Abstract
|
C00143 I3675 00 |
Monte Carlo Radiation Transport Code Systems for Fallout Shielding. |
GRENADE
|
Abstract
|
C00516 C1787 00 |
Green's Function Nodal Algorithm for the Diffusion Equation. |
GRENADE
|
Abstract
|
C00516 D0780 00 |
Green's Function Nodal Algorithm for the Diffusion Equation. |
GRESS 3.0 |
Abstract
|
P00231 MFMWS 02 |
Gradient Enhanced Software System. |
GRETEL |
Abstract
|
P00100 I0370 00 |
Analyzer and Processor of Ge(Li) Gamma-Ray Spectrometric Data. |
GRFPAK |
Abstract
|
P00478 I0360 00 |
Code System to Plot CORTES FEM Results. |
GROUP STRUCTURE |
Abstract
|
D00156 ALLCP 00 |
Standard Energy Group Structures Of Cross Section Libraries For Reactor Shielding, Reactor Cell Fusion Neutronics Applications: VITAMIN-J, ECC0-33, ECC0-2000. |
GROUPXS |
Abstract
|
P00246 C0740 00 |
Processing of Double-Differential Cross Sections in the New ENDF-VI Format. |
GRPANL |
Abstract
|
P00321 D0VAX 00 |
Code System for Analyzing Ge and Alpha-Particle Detector Spectra. |
GRTUNCL3D
|
Abstract
|
C00721 MNYCP 01 |
Code to Calculate Semi-Analytic First Collision Source and Uncollided Flux. |
GT2R2 |
Abstract
|
P00483 ALLMF 00 |
Code System to Calculate Fuel Rod Thermal Performance. |
GUI2QAD-3D
|
Abstract
|
C00697 PC586 01 |
Point Kernel Code System for Neutron and Gamma-Ray Shielding Calculations in Complex Geometry, Including a Graphical User Interface. |
HAARM-3 |
Abstract
|
P00401 CDCMF 00 |
Aerosol Behavior Log-Normal Distribution Model. |
HABIT 1.1
|
Abstract
|
C00665 IBMPC 01 |
Code System for Evaluation of Control Room Habitability. |
HADOC
|
Abstract
|
C00452 U1100 00 |
Calculates External and Inhalation Doses from Acute Radionuclide Releases on the Hanford Site. |
HALLMARK |
Abstract
|
D00005 I0360 00 |
Discrete Ordinates and Monte Carlo Results of Neutron and Secondary Gamma-Ray Transport in Air-Over-Ground Geometry. |
HAM
|
Abstract
|
C00267 U1108 00 |
Monte Carlo Multigroup Neutron and Photon High Altitude Transport Code System. |
HARAD
|
Abstract
|
C00387 I0360 00 |
Calculation of Daughter Concentrations in Air Following the Atmospheric Release of a Parent Radionuclide. |
HATCHES-12 |
Abstract
|
D00206 PC486 00 |
Thermodynamic Database for Radiochemical Modelling. |
HAUSER*5 |
Abstract
|
P00152 U0000 00 |
Code System for Calculating Nuclear Cross Sections. |
HEATING 7.3 |
Abstract
|
P00199 MNYCP 06 |
Multidimensional, Finite-Difference Heat Conduction Analysis Code System, Versions 7.2i and 7.3. |
HECTR 1.5+
USSO
|
Abstract
|
P00457 CY000 00 |
Hydrogen Event Containment Response Code System. |
HEITLER |
Abstract
|
P00004 I7030 00 |
Cross Section Generator. |
HELLO |
Abstract
|
D00058 I0360 00 |
47 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 60 MeV. |
HERAD
|
Abstract
|
C00444 CY00I 00 |
Three-Dimensional Monte Carlo Computer Code System for Calculating Radiation Damage from Ion Beams. |
HERMES-KFA
|
Abstract
|
C00687 MNYWS 00 |
Monte Carlo Code System for High-Energy Radiation Transport Calculations. |
HEXAB-3D
|
Abstract
|
C00593 I0370 00 |
Three-Dimensional Few-Group Coarse Mesh Diffusion Code for Neutron Physics Calculation of Reactor Core in Hexagonal Geometry. |
HIC-1
|
Abstract
|
C00249 I0360 00 |
Monte Carlo Code System for Calculating Heavy Ion Reactions at Energies > 50 MeV/Nucleon. |
HILO |
Abstract
|
D00087 I0370 00 |
66 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 400 MeV. |
HILO2K |
Abstract
|
D00220 MNYCP 00 |
Coupled 83 Neutron, 22 Photon Group Cross Sections for Neutron Energies Up to 2 GeV. |
HILO86 |
Abstract
|
D00119 I0360 00 |
66 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV. |
HILO86 |
Abstract
|
D00119 PC386 01 |
66 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV. |
HILO86R |
Abstract
|
D00187 ALLCP 00 |
66 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV. |
HIMAC
| Abstract
|
M00001 MNYCP 02 |
Experimental Data of Neutron Yields from Thick Targets Bombarded by 100 to 800 MeV / Nucleon Heavy Ions. |
HORN
|
Abstract
|
C00568 I3083 00 |
A Computer Code To Analyze The Gas-Phase Transport of Fission Products In Reactor Cooling System Under Severe Accidents. |
HOTSPOT 2.05
|
Abstract
|
C00644 IBMPC 03 |
Health Physics Code System for Evaluating Accidents Involving Radioactive Materials. |
HPICE |
Abstract
|
D00007 I0360 05 |
Evaluated Photon Interaction Library, ENDF/B File 23 Format. |
HPPOS 1.5 |
Abstract
|
D00173 IBMPC 00 |
Health Physics Position Database. |
HPPOS V2 |
Abstract
|
D00173 IBMPC 01 |
Health Physics Positions (HPPOS) Data Base Based on Current 10 CFR 20. |
HSI-DRG |
Abstract
|
P00435 IBMPC 00 |
Code System for Use with Human System Interface Design Review Guidelines. |
HUGO |
Abstract
|
D00099 I3033 00 |
Photon Interaction Data in ENDF/B-V Format. |
HUGO VI |
Abstract
|
D00146 I3033 00 |
Photon Interaction Data in ENDF/B-VI Format. PHOTB6 in DLC-179/ENDLIB-97 is an updated version of these data. |
HYACINTH
|
Abstract
|
C00294 I0360 00 |
Fast Heavy Isotope Point Burnup and Decay Code System - Analytical Solution. |
HYPERMET |
Abstract
|
P00101 C3800 00 |
Gamma-Ray Spectra Analyzer Germanium Detector. |
HYPERMET |
Abstract
|
P00101 F150F 00 |
Gamma-Ray Spectra Analyzer Germanium Detector. |
HYPERMET |
Abstract
|
P00101 I0360 00 |
Gamma-Ray Spectra Analyzer Germanium Detector. |
I-R-MAN |
Abstract
|
D00050 ALLCP 00 |
Photon Interaction Data on ICRP Reference Man. |
ICAR |
Abstract
|
P00291 IPCAT 00 |
A Code For Combinatorial Calculation of Level Densities. |
ICOM
|
Abstract
|
C00651 PC386 00 |
Code System for Calculating Ion Track Condensed Collision Model. |
IDC
|
Abstract
|
C00384 I0360 00 |
ICRP Dosimetric Calculational System. |
IEAF-2001 |
Abstract
|
D00217 MNYCP 00 |
Intermediate Energy Activation File - 2001. |
IER |
Abstract
|
P00024 I3675 00 |
A Gauss-based Quadrature Formula Applied to Sievert's Integral. An Exponential Integral Routine. |
IMPACTS-BRC2.1
|
Abstract
|
C00666 IBMPC 00 |
Code System for Analysis of Potential Radiological Impacts. |
IMPORTANCE |
Abstract
|
P00407 I0370 00 |
FTA Basic Event & Cut Set Ranking. |
INAP
|
Abstract
|
C00235 U1108 00 |
Improved Neutron Activation Prediction Code Systems. |
INDOS
|
Abstract
|
C00236 DP010 00 |
Conversational Computer Code Systems to Implement ICRP-10-10A Models for Estimation of Internal Radiation Dose to Man. |
INDOSE V2.1.1
|
Abstract
|
C00720 PC586 00 |
Internal Dosimetry Code System Using Biokinetics Models |
INDRA
|
Abstract
|
C00303 I0360 00 |
A Modular System for Calculating the Neutronics and Photonics Characteristics of a Fusion Reactor Blanket. |
INFLTB |
Abstract
|
P00313 ALLCP 00 |
Gamma-Ray Absorption Coefficient Calculation. |
INGDOS
|
Abstract
|
C00408 DP010 00 |
A Conversational Code System Designed to Implement NRC Reg-Guide 1.109 Models for Estimation of Annual Doses from Ingestion of Atmospherically Released Radionuclides in Foods. |
INGEN |
Abstract
|
P00207 C0000 00 |
A General-Purpose Mesh Generator for Finite Element Codes. |
INREM II
|
Abstract
|
C00392 I3033 00 |
Computer Implementation of Recent Models for Estimating the Dose Equivalent to Organs of Man from an Inhaled or Ingested Radionuclide. |
INREM/EXREM
|
Abstract
|
C00185 I0360 00 |
Beta and Gamma Radiation Environmental Dose Code Systems. |
INTERTRAN I
|
Abstract
|
C00473 ALLMF 00 |
A Code System for Assessing the Impact from Transporting Radioactive Material. |
INTRIGUE-II |
Abstract
|
P00054 I0360 00 |
Logarithmic and Semilogarithmic CALCOMP Plot Routines. |
INTRUDE-ANS
|
Abstract
|
C00539 D8810 00 |
A Repository Intrusion Risk Evaluation Code. |
INVENT
|
Abstract
|
C00540 D8810 00 |
A Radionuclide Inventory and Hazard Index Code. |
IODES
|
Abstract
|
C00365 I0360 00 |
A Code System for Calculating the Estimation of Dose to the World Population from Releases of Iodine-129 to the Environment. |
IONMIG
|
Abstract
|
C00526 ALLMF 00 |
Code System for Radionuclide Migration Calculations. |
IRAN-LIB |
Abstract
|
D00159 IBMPC 00 |
A P-3 Coupled Neutron-Gamma Cross Section Library in ISOTXS For Use with ANISN/PC (CCC-514). |
IRDAM
|
Abstract
|
C00524 IPCXT 00 |
Interactive Rapid Dose Assessment Model. |
IRDF-2002 |
Abstract
|
D00229 MNYCP 01 |
The International Reactor Dosimetry File. |
IRDF-90 |
Abstract
|
D00161 ALLCP 01 |
The International Reactor Dosimetry File. |
IRDF82 |
Abstract
|
D00094 I0360 00 |
International Reactor Dosimetry Data. |
IRRAS 4.16
USSO
|
Abstract
|
P00386 IBMPC 04 |
Code System to Calculate Integrated Reliability and Risk Analysis. |
ISO-PC 2.1
|
Abstract
|
C00636 IBMPC 01 |
Kernel Integration Code System for General Purpose Isotope Shielding Analyses. |
ISOGEN II
|
Abstract
|
C00055 I3675 00 |
Radioisotope Generator Code. |
ITER-2 |
Abstract
|
P00148 C0000 00 |
Codes for Unfolding Activation Detector Data and Pulse Height Spectra. |
ITS 3.0
|
Abstract
|
C00467 MNYCP 02 |
Integrated TIGER Series of Coupled Electron/Photon Monte Carlo Transport Codes System. |
JDL-IMPORTANCE
| Abstract
|
M00005 MNYCP 00 |
Adjoint Function: Physical Basis of Variational & Perturbation Theory in Transport & Diffusion Problems. |
JDL-REACTOR-KIN
| Abstract
|
M00006 MNYCP 00 |
Nuclear Reactor Kinetics and Control. |
JDL-THERMODYNAM
| Abstract
|
M00007 MNYCP 00 |
Thermodynamics: Frontiers and Foundations. |
JENDL-2 |
Abstract
|
D00122 FM380 00 |
Japanese Evaluated Neutron Cross Section Data in ENDF/B-IV Format. |
JENDL/D-99 |
Abstract
|
D00204 MNYCP 00 |
JENDL Dosimetry File 99. |
JFS |
Abstract
|
D00070 ALLCP 00 |
Japanese Evaluated Nuclear Data Library. |
JFS |
Abstract
|
D00111 I3033 00 |
70 Group Neutron Fast Reactor Cross Section Set and 25 Group Neutron Fast Reactor Cross Section Set. |
JFS3J2 |
Abstract
|
D00108 FM200 00 |
70 Group Neutron Fast Reactor Cross Section Set Based on JENDL-2B. |
JIMCOF |
Abstract
|
D00078 F2307 00 |
Multigroup Constants fFle Based on ENDF/B IV. |
JN-METD 2&1
|
Abstract
|
C00208 I0370 00 |
Neutron Transport Code System with Isotropic Scattering, Bare Slabs and Homogeneous Slabs (JN Method 1), Multilayer Slabs (JN Method 2). |
K009
|
Abstract
|
C00062 I7090 00 |
Solid Angle Integration Charged Particle Penetration Code. |
K019
|
Abstract
|
C00100 I0360 00 |
Shield Thickness Calculation Program for Space Vehicles. |
KAMCCO
|
Abstract
|
C00325 I0370 00 |
Three-Dimensional Time Dependent Monte Carlo Code System for Fast Neutron Physics Problems. |
KAOS-V |
Abstract
|
P00306 CY000 00 |
An Evaluation Tool For Neutron Kerma Factors and Other Nuclear Responses. |
KAOS/LIB-V |
Abstract
|
D00160 CY000 00 |
A Library of Nuclear Response Functions Generated by KAOS-V Code From ENDF/B-V and Other Data Files. |
KAP-VI
|
Abstract
|
C00094 U1108 00 |
Kernel Integration Code System in Complex Geometry. |
KDDK |
Abstract
|
D00061 I0360 00 |
Measured Results of Delayed Beta- and Gamma-Ray Spectra due to Thermal-Neutron Fission of U-235. |
KDLIBE
|
Abstract
|
C00124 I3675 00 |
Kernel-Diffusion Shielding Analysis System. |
KEDAK3 |
Abstract
|
D00141 I0370 00 |
Evaluated Neutron Nuclear Data for Reactor Physics Calculations. |
KENO2MCNP |
Abstract
|
P00541 PC586 00 |
Conversion of Input Data between KENO V.a and MCNP File Formats, Version 5L. |
KERMAL |
Abstract
|
D00142 ALLCP 00 |
Neutron and Gamma-Ray Kerma Factors Based on LLNL Nuclear Data Files. |
KERNEL
|
Abstract
|
C00672 IBMPC 00 |
Monte Carlo Code System for Electron (Positron) Dose Kernel Calculations. |
KFIX |
Abstract
|
P00409 C7600 00 |
Code System to Calculate Transient 2-Dimensional 2-Fluid Flow Dynamics. |
KFIX 3D |
Abstract
|
P00383 C7600 00 |
Code System to Calculate Three-Dimensional Extension Two-Phase Flow Dynamics. |
KIM
|
Abstract
|
C00376 I3033 00 |
A Two-Dimensional Monte Carlo Code System for Linear Neutron Transport Calculations. |
KORIGEN
|
Abstract
|
C00457 I3033 00 |
A Modification of the Isotope Generation and Depletion Code System ORIGEN. CCC-702/ORIGEN-ARP is recommended for new ORIGEN users. |
KRONIC
|
Abstract
|
C00229 I0360 00 |
Calculation of Annual Average External (Beta and Gamma Radiation) Doses from Chronic Atmospheric Releases of Radionuclides. |
KRONIC
|
Abstract
|
C00229 U1108 00 |
Calculation of Annual Average External (Beta and Gamma Radiation) Doses from Chronic Atmospheric Releases of Radionuclides. |
KUX
|
Abstract
|
C00515 ALLCP 00 |
Medical X-Ray Shielding Calculation. |
KX-RAY |
Abstract
|
D00021 I0360 00 |
Evaluated X-ray Cross Section Library. |
L26P3S34 |
Abstract
|
D00112 IBMMF 00 |
ENDL 26-Group up to P3 Library Prepared by SUPERTOG for 34 Materials. |
LA100 |
Abstract
|
D00168 ALLCP 00 |
Evaluated Nuclear Data Library for Transport Calculations Involving Incident Neutrons and Protons of Energy Up to 100 MeV. |
LABAN-PEL
|
Abstract
|
C00611 IMFPC 00 |
A Two-Dimensional, Multigroup Diffusion, High-Order Response Matrix Code. |
LADTAP II
|
Abstract
|
C00363 C7600 00 |
Code System for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents. |
LADTAP II
|
Abstract
|
C00363 D0780 00 |
Code System for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents. |
LADTAP II
|
Abstract
|
C00363 I3033 00 |
Code System for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents. |
LAFPX-V |
Abstract
|
D00054 C0000 01 |
A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections. |
LAFPX-V |
Abstract
|
D00054 C0000 02 |
A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections. |
LAHET 2.8
|
Abstract
|
C00696 MFMWS 00 |
Code System for High Energy Particle Transport Calculations. |
LAHIMACK |
Abstract
|
D00128 I0360 00 |
A Multigroup Library of Neutron and Gamma Cross Sections and Response Functions in the Energy Range up to 800 MeV. |
LAPHANO |
Abstract
|
P00020 C6600 00 |
PO Multigroup Photon Production Matrix and Source Vector Code for ENDF Data. |
LAPHANO |
Abstract
|
P00020 I0360 00 |
PO Multigroup Photon Production Matrix and Source Vector Code for ENDF Data. |
LAPUR6
USSO
|
Abstract
|
P00395 PC586 02 |
BWR Core Stability Measurements. |
LAS CRUCES
USSO
|
Abstract
|
D00194 ALLCP 00 |
Las Cruces Trench Site Database, Vadose Model. |
LASER
|
Abstract
|
C00344 I0360 00 |
A One-Dimensional, Neutron-Thermalization, Lattice-Cell Program Based on MUFT and THERMOS. |
LEAF
|
Abstract
|
C00312 C6600 00 |
Fission Product Release Calculator-From a Reactor Containment Building for Arbitrary Radioactive Decay Chains. |
LEAP-ADDELT |
Abstract
|
P00138 I0360 00 |
Multigroup Thermal Neutron Scattering Data Generator for Hydrogen in Light Water and Deuterium in Heavy Water. |
LEBC
|
Abstract
|
C00052 I7090 00 |
Electron Bremsstrahlung Code. |
LEGENDRE FUNCTI |
Abstract
|
P00108 I0360 00 |
Legendre Functions of the First Kind and Legendre Polynomials. |
LENDL |
Abstract
|
D00034 I0360 02 |
Livermore Evaluated Neutron and Secondary Gamma-Ray Production Cross-Section Library in ENDF/B-IV Format. |
LENDL V |
Abstract
|
D00120 I0360 00 |
Lawrence Livermore National Laboratory Evaluated Nuclear Data Library in ENDF-V Format. |
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. |
LEP |
Abstract
|
D00001 I0360 02 |
Cascade and Evaporation Particle Results from Low-Energy Intranuclear Cascade Calculations. |
LEPRICON |
Abstract
|
P00277 IRISC 00 |
PWR Pressure Vessel Surveillance Dosimetry Analysis System. |
LEPRICON |
Abstract
|
P00277 I3033 01 |
PWR Pressure Vessel Surveillance Dosimetry Analysis System. |
LG-H
|
Abstract
|
C00087 I7090 00 |
Ray Analysis Cylindrical Duct Kernel Code for Neutrons and Gamma Rays. |
LGH-G
|
Abstract
|
C00239 I0360 00 |
Calculation of Gamma Radiation through Partially Shielded Gaps (Buildup Factor Method in Taylors Approximation). |
LHS |
Abstract
|
P00394 PC386 00 |
Code System to Generate Latin Hypercube and Random Samples. |
LHS |
Abstract
|
P00394 SUN05 00 |
Code System to Generate Latin Hypercube and Random Samples. |
LIB123 |
Abstract
|
D00153 ALLCP 00 |
AMPX-II P3 123-Group Neutron Cross Section Master Interface Library. |
LIBMAK |
Abstract
|
P00087 I0360 00 |
ANISN-Type Binary Data Processing Code System. |
LINEDOSE
|
Abstract
|
C00468 IBMPC 00 |
A Line Source Shielding Code for Personal Computers. |
LINSED
|
Abstract
|
C00673 I0360 00 |
1D Multireach Sediment Transport Model |
LIONS
|
Abstract
|
C00247 I0360 00 |
Calculation of Fission Product Inventory, Gamma-Ray Dose Rates and Gamma-Ray Doses by Kernel Integration. |
LOGNORML |
Abstract
|
P00307 IPCAT 00 |
Lognormal Probability Analysis Code System for Estimating Doses in Epidemiologic Studies. |
LOOM-P |
Abstract
|
P00153 F2307 00 |
A Finite Element Mesh Generation Code System with On-Line Graphic Display. |
LOUHI82 |
Abstract
|
P00236 U1108 00 |
General Purpose Unfolding Program with Linear and Nonlinear Regularizations. |
LPGS
|
Abstract
|
C00385 I3033 00 |
Code System for Calculating Radiation Exposure Resulting from Accidental Radioactive Releases to the Hydrosphere. |
LPPC
|
Abstract
|
C00051 I7090 00 |
Proton Penetration Code. |
LPSC
|
Abstract
|
C00064 I7090 00 |
Proton Penetration Code - Multilayer Slab Geometry. |
LPTAU |
Abstract
|
P00340 MNYCP 00 |
Quasi-Random Sequence Generators. |
LRSPC
|
Abstract
|
C00050 I7090 00 |
Range and Stopping Power Calculator. |
LSHINSE
|
Abstract
|
C00554 IBMPC 00 |
Calculates Flux and Dose Rate from the Scattering of Radiation in Air. |
LSL-M2 |
Abstract
|
P00233 D6220 00 |
Least-Squares Logarithmic Adjustment of Neutron Spectra. |
LSL-M2 |
Abstract
|
P00233 IBMPC 00 |
Least-Squares Logarithmic Adjustment of Neutron Spectra. |
LSMOD-GLSMOD |
Abstract
|
P00342 IBMPC 00 |
A Least-Squares Computational Tool Kit. |
LSVDC
|
Abstract
|
C00053 I7090 00 |
Space Vehicle Dose Calculation. |
LSVDC
|
Abstract
|
C00053 I7090 01 |
Space Vehicle Dose Calculation. |
LTC |
Abstract
|
P00329 IBMPC 00 |
LMR Transient Calculation Code System (version 5). |
LUIN-II
|
Abstract
|
C00220 C6600 00 |
Analytical Straight-Ahead Transport Code System-Calculation of Cosmic-Ray Spectra, Fluxes and Ionization in the Earth's Atmosphere. |
LUMP |
Abstract
|
D00089 I0360 00 |
Evaluated Lumped Fission Product Cross Sections for Fast Reactor Analysis--Based on ENDF/B-V Data. |
MACK-IV |
Abstract
|
P00132 I3691 00 |
Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format. |
MACKLIB |
Abstract
|
D00029 I3675 00 |
100 Group Neutron Kerma Factors and Reaction Cross Sections Generated by MACK from Data in ENDF Format. |
MACKLIB-IV-82 |
Abstract
|
D00060 I0360 01 |
A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV. |
MADONNA
|
Abstract
|
C00425 I0370 00 |
Two-dimensional Neutron Streaming Coupled Removal-Diffusion-Albedo-Transport Code System. |
MAEROS |
Abstract
|
P00466 C7600 00 |
Code System for Multicomponent Aerosol Time Evolution. |
MAGIK
|
Abstract
|
C00359 I0360 00 |
A Monte Carlo Code System for Computing Induced Residual Activation Dose Rates. |
MAGNA
|
Abstract
|
C00158 C3600 00 |
Multi-Source Gamma-Ray Kernel Integration Code System. |
MAINTAIN |
Abstract
|
P00067 I0360 00 |
Code System for Use in Maintaining and Revising Card Image Files on Tape. |
MANYFILE |
Abstract
|
P00068 I0360 00 |
Utility Routine - Manipulation of Data Sets Between Various I-O Devices. |
MAP
|
Abstract
|
C00150 I3675 00 |
Kernel Integration Code System in Complex Geometry with Special Application to Surface Sources Determined by Discrete Ordinates Calculations. |
MARC-PN
|
Abstract
|
C00311 D8810 00 |
A Neutron Diffusion Code System with Spherical Harmonics Option. |
MARC-PN
|
Abstract
|
C00311 I3081 00 |
A Neutron Diffusion Code System with Spherical Harmonics Option. |
MARCH2 |
Abstract
|
P00473 CDCMF 00 |
Code System to Model LWR Meltdown Accident Response. |
MARCOPOLO |
Abstract
|
P00225 I0360 00 |
Code System for Calculating the Radial and Axial Neutron Diffusion Coefficients in One-Group and Multigroup Theory. |
MARD 4.16 |
Abstract
|
P00448 IBMPC 00 |
Models And Results Database System. |
MARIA SYSTEM |
Abstract
|
P00359 D6000 00 |
Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations. |
MARINRAD
|
Abstract
|
C00503 C1785 00 |
Code System Model for Assessing the Consequences of Release of Radioactive Material into the Oceans. |
MARLOWE 15B |
Abstract
|
P00137 MNYCP 08 |
Computer Simulation of Atomic Collisions in Crystalline Solids (Version 15). |
MARMER
|
Abstract
|
C00579 D8350 00 |
A Flexible Point-Kernel Shielding Code System. |
MARMER
|
Abstract
|
C00579 PC486 00 |
Flexible Point-Kernal Shielding Code System. |
MARS |
Abstract
|
P00117 I0360 00 |
Collection of Computer Codes for Manipulating Multigroup Cross Section Libraries in AMPX or CCCC Formats. |
MARTHA |
Abstract
|
P00232 I0360 00 |
Monte Carlo Response Function Calculation for Sodium Iodide Photon Detectors. |
MASS |
Abstract
|
D00025 I0360 01 |
Atomic Mass Evaluation. |
MATADOR
|
Abstract
|
C00689 CDCMF 00 |
Radionuclide Behavior in Containments. |
MATEXP |
Abstract
|
P00059 I0360 00 |
Matrix Exponential Method Applied to Systems of Ordinary Differential Equations. |
MATXS1 |
Abstract
|
D00114 C0000 00 |
30-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-IV in MATXS Format. |
MATXS10 |
Abstract
|
D00176 ALLCP 00 |
30-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-VI in MATXS Format. |
MATXS11 |
Abstract
|
D00177 ALLCP 00 |
80-Group Neutron, 24-Group Photon Cross Sections from ENDF/B-VI in MATXS Format. |
MATXS175/42-JE |
Abstract
|
D00151 D8810 00 |
JEF/EFF Based VITAMIN-J 175 Neutron, 42 Photon Multigroup Data Library in MATXS Format. |
MATXS5A |
Abstract
|
D00115 C0000 00 |
30-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-V in MATSX Format. |
MATXS6A |
Abstract
|
D00116 C0000 00 |
80-Group Neutron, 24-Group Photon Fast-Reactor Cross Section from ENDF/B-V in MATXS Format. |
MATXS70-JEF87 |
Abstract
|
D00148 D8810 00 |
JEF/EFF Based 70 Group Neutron Data Library in MATXS Format. |
MATXS7A |
Abstract
|
D00117 C0000 00 |
69-Group Thermal-Reactor Neutron Cross Section Data from ENDF/B-V in MATXS Format. |
MATXUF |
Abstract
|
P00130 I0360 00 |
On-Line Derivative Method, Spectrum Unfolding Code System for NE-213 Liquid Fast Scintillation Proton Recoil Data. |
MAVRAC
|
Abstract
|
C00023 I7090 00 |
Model Astronaut and Vehicle Radiation Analysis Code. |
MAX-XTREME |
Abstract
|
P00001 C0000 00 |
Generalized Several-Constraint LaGrange Multiplier. |
MAZE II |
Abstract
|
P00041 U1108 00 |
Spectral Unfolding Code. |
MAZE-1 |
Abstract
|
P00041 C6600 00 |
Spectral Unfolding Code. |
MC**2-2 |
Abstract
|
P00350 SUN05 01 |
Code System for Calculating Fast Neutron Spectra and Multigroup Cross-sections from ENDF/B Data (November 2000 Version). |
MCB1C
|
Abstract
|
C00719 MNYWS 00 |
Monte-Carlo Continuous Energy Burnup Code System. |
MCB63NEA.BOLIB |
Abstract
|
D00216 MNYCP 00 |
ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code. |
MCFLARE
|
Abstract
|
C00093 I7090 00 |
Monte Carlo Code to Simulate Solar Flare Events and Estimate Probable Doses Encountered on Interplanetary Missions. |
MCJEF22NEA.BOLIB |
Abstract
|
D00203 MNYCP 01 |
JEF 2.2 Cross Section Library for the MCNP Monte Carlo Code. |
MCJEFF3.1NEA |
Abstract
|
D00228 MNYCP 00 |
Neutron Cross Section Library Based on JEFF3.1 for Use with MCNP. |
MCNP-DSP
|
Abstract
|
C00699 MNYCP 00 |
Monte Carlo N-Particle Transport Code System with Digital Signal Processing based on MCNP4A. |
MCNP-POLIMI
|
Abstract
|
C00718 PC586 00 |
Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities Based on MCNP4C. |
MCNP5/MCNPX
|
Abstract
|
C00740 MNYCP 02 |
Monte Carlo N-Particle Transport Code System Including MCNP5 1.51 and MCNPX 2.6.0 and Data Libraries (Source & Executables). |
MCNP5/MCNPX-EXE
|
Abstract
|
C00740 MNYCP 03 |
Monte Carlo N-Particle Transport Code System Including MCNP5 1.51 and MCNPX 2.6.0 and Data Libraries (Executables - No Source). |
MCNPDATA |
Abstract
|
D00200 ALLCP 03 |
Standard Neutron, Photon, and Electron Data Libraries for MCNP4C or MCNP-PoliMi. |
MCNPXS |
Abstract
|
D00189 ALLCP 00 |
Standard Neutron, Photon, and Electron Data Libraries for MCNP4B or MCNP-DSP. |
MCRAC
|
Abstract
|
C00562 IBMPC 00 |
Multiple Cycle Reactor Analysis Code. |
MCRTOF
|
Abstract
|
C00435 FM200 00 |
Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region. |
MCRTOF
|
Abstract
|
C00435 I0360 00 |
Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region. |
MCVIEW |
Abstract
|
P00202 FM780 00 |
View Factor Calculation for Three-Dimensional Geometries. |
MECC-7
|
Abstract
|
C00156 I0360 00 |
Medium-Energy Intranuclear Cascade Code System. |
MEDUSA-IB
|
Abstract
|
C00505 HM200 00 |
One-Dimensional Lagrangian Code for Plasma Hydrodynamic Analysis of a Fusion Pellet Driven by Ion Beams. |
MEDUSA-PIJ
|
Abstract
|
C00349 F2307 00 |
One-Dimensional Laser Fusion Analyzer (Including Neutron Heating Effect) Collision Probability Method. |
MENDL-2P |
Abstract
|
D00207 MNYCP 00 |
Proton Reaction Data Library for Nuclear Activation (Medium Energy Nuclear Data Library.) |
MENSLIB |
Abstract
|
D00084 I0370 00 |
60 Group, P5, Cross Sections in DTF-IV for Transport Calculations for Neutrons with Energies Up to 60 MeV. |
MERCURE 4-82
|
Abstract
|
C00142 I3033 00 |
Three-Dimensional Code System for Integrating Multigroup Line-of-Sight Attenuation Kernels by Monte Carlo Techniques. |
MESA |
Abstract
|
P00223 I3033 00 |
Non-Linear Least Squares Spectral Analysis. |
MESODIF-II
|
Abstract
|
C00498 D0780 00 |
A Variable Trajectory Plume Segment Model to Assess Ground-Level Air Concentrations and Depositions of Routine Effluent Releases from Nuclear Power Facilities. |
MESOI
|
Abstract
|
C00497 D0780 00 |
Interactive Mesoscale Lagrangian Puff Dispersion Model with Deposition and Decay. See CCC-677/MESORAD. |
MESORAD 1.4
|
Abstract
|
C00677 D0VAX 00 |
Code System for Emergency Response Dose Assessment. |
MESYST
|
Abstract
|
C00706 MNYWS 00 |
Code System to Simulate 3D Tracer Dispersion in Atmosphere. |
METD |
Abstract
|
P00197 DGMV1 00 |
Computer Code Systems for Use with Meteorological Data. |
METD |
Abstract
|
P00197 I3033 00 |
Computer Code Systems for Use with Meteorological Data. |
MEVDP
|
Abstract
|
C00157 C6600 00 |
Primary Radiation Transport Code System - Complex Geometry - Computerized Anatomical Model Man. |
MGA8 |
Abstract
|
P00542 MNYCP 00 |
Code System to Determine Pu Isotope Abundances from Multichannel Analyzer Gamma Spectra. |
MGCLIB |
Abstract
|
D00118 FM380 00 |
137 and 26 Neutron Multigroup Cross Section Library with the Bondarenko Type Shielding Table. |
MICAP |
Abstract
|
P00261 I3033 00 |
A Monte Carlo Code System for Analysis of Ionization Chamber Responses. |
MICROX-2 |
Abstract
|
P00374 MNYCP 02 |
Code System to Create Broad-Group Cross Sections with Resonance Interference and Self-Shielding from Fine-Group and Pointwise Cross Sections. |
MIGROS3 |
Abstract
|
P00265 I0370 00 |
A Code for the Generation of Group Constants for Reactor Calculations from Neutron Nuclear Data in KEDAK Format. |
MILDOS
|
Abstract
|
C00398 C0000 00 |
Calculation of Radiation Doses from Uranium Recovery Operations. |
MILDOS-AREA
|
Abstract
|
C00608 IBMPC 00 |
Calculation of Radiation Dose from Uranium Recovery Operations for Large-Area Sources. |
MINET |
Abstract
|
P00490 CY000 00 |
Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis. |
MINIGAL |
Abstract
|
P00180 I3033 00 |
Neutron Cross Section Processing System for Calculating Average Values from Data in the Standard United Kingdom Nuclear Data Library Format. |
MINTEQ |
Abstract
|
P00494 DVX11 00 |
Code System to Model Aqueous Geochemical Equilibria. |
MINX |
Abstract
|
P00105 C6600 00 |
Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats. |
MINX |
Abstract
|
P00105 I0360 00 |
Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats. |
MISSIONARY |
Abstract
|
P00114 I0360 00 |
ENDF/B to NDL Data Format Converter. |
MIXEN |
Abstract
|
P00318 IRISC 00 |
Code System to Replace Files 4 and 6 of ENDF-6 with Files 4 and 5 of ENDF/B-IV. |
MKENO-DAR
|
Abstract
|
C00513 FM380 00 |
Direct Angular Representation Monte Carlo Code for Criticality Safety Analysis |
MMCR
|
Abstract
|
C00441 FM200 00 |
Multigroup Monte Carlo Neutron and Photon Transport Code. |
MOCA
|
Abstract
|
C00590 IPCAT 00 |
Monte Carlo Criticality Code System for Hexagonal Geometries. |
MOCUP |
Abstract
|
P00365 DALPU 00 |
MCNP/ORIGEN Coupling Utility Programs. |
MODEL
|
Abstract
|
C00329 I3033 00 |
Models of Trapped Proton and Electron Environments for Solar Maximum and Minimum. |
MOMENT I
|
Abstract
|
C00188 U1108 00 |
Moments Method Neutron Transport Code System. |
MOMGEM-MOMDIS
|
Abstract
|
C00085 I7090 00 |
Moments Method Reconstruction of Scattered Gamma-Ray Distributions. |
MONK 6.3
FEDC
|
Abstract
|
C00393 I3033 00 |
A General Purpose Monte Carlo Neutronics Code System. |
MONTEBURNS 2.0 |
Abstract
|
P00455 MNYCP 02 |
An Automated, Multi-Step Monte Carlo Burnup Code System. |
MONTUK-80 |
Abstract
|
D00072 ALLCP 01 |
UKCTR III Transmutation and Activation Data, 100-Group Neutron Activation Cross-Section Data for Fusion Reactor Structure and Coolant Materials. |
MORECA |
Abstract
|
P00411 PC386 00 |
Computer Code System for Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup. |
MORN |
Abstract
|
P00062 I0360 00 |
Calculation of the Response of Sodium Iodide Crystals to Gamma Rays. |
MORSE-ALB
|
Abstract
|
C00394 FM200 00 |
Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System, Albedo Version. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-ANSI STD.
|
Abstract
|
C00127 I3675 00 |
A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-B
|
Abstract
|
C00368 I0370 00 |
General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-C
|
Abstract
|
C00431 C7600 00 |
Monte Carlo Multigroup Neutron Code System for the Solution of Criticality Problems. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-CG
|
Abstract
|
C00203 C0000 00 |
A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-CG
|
Abstract
|
C00203 CY000 00 |
A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-CG
|
Abstract
|
C00203 D0VAX 00 |
A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-CG
|
Abstract
|
C00203 I0360 00 |
A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-CG
|
Abstract
|
C00203 U0000 00 |
A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-CGA
|
Abstract
|
C00474 ALLCP 03 |
A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Array Geometry Capability, Version 2. |
MORSE-CV
|
Abstract
|
C00535 HM280 00 |
Multigroup Neutron and Gamma-Ray Monte Carlo Transport Code with Covariance Calculation. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-E
|
Abstract
|
C00258 I0360 00 |
Special Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-EMP
|
Abstract
|
C00588 IBMPC 00 |
General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Array Geometry Capability. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-H
|
Abstract
|
C00471 I3081 00 |
A Revised Version of the MORSE Monte Carlo Radiation Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-L
|
Abstract
|
C00261 C6600 00 |
Multigroup Neutron and Gamma-Ray Transport Code System for the Solution of Penetration Problems. We recommend C00474/ALLCP/02 MORSE-CGA. |
MORSE-SGC
|
Abstract
|
C00277 C7600 00 |
A Super Grouped Cross Section Version of the MORSE Code System. We recommend either C00474/ALLCP/02 MORSE-CGA, or C00545/IRISC/01 SCALE 4.2. |
MORSEC-SP2 |
Abstract
|
P00142 H6000 00 |
A Multigroup Cross Section Module for the MORSE Monte Carlo Computer Code System. |
MOSRA-LIGHT |
Abstract
|
P00505 MNYWS 00 |
High-Speed Three-Dimensional Nodal Diffusion Code System. |
MOXY-MOD32 |
Abstract
|
P00385 I0360 00 |
BWR Core Heat Transfer Code System. |
MRIPP 1.0
|
Abstract
|
C00655 PC386 00 |
Magnetic Resonance Image Phantom Code System to Calibrate in vivo Measurement Systems. |
MRSPAK |
Abstract
|
P00212 DVX11 00 |
A Code System To Generate a Text File Containing Combinatorial Geometry Data Corresponding to PADL2 Geometry. |
MSM-SOURCE |
Abstract
|
P00369 MNYCP 00 |
Code System for Generation of Input Data for MCNP. |
MTR_PC 2.6
|
Abstract
|
C00674 PC386 00 |
Modular Code System for Neutronics, Thermalhydraulics and Shielding Calculations. |
MULTI-KENO2
|
Abstract
|
C00492 FM380 00 |
A Monte Carlo Code System for Criticality Safety Analysis. |
MUP2 |
Abstract
|
P00289 I3090 00 |
A Program to Calculate Fast Neutron Data for Medium-Heavy Nuclei. |
MURLI
|
Abstract
|
C00378 DP011 00 |
Integral Transport Theory Code System for Thermal Reactor Lattice Cell Calculation. |
MUSCAT
|
Abstract
|
C00281 I0360 00 |
Calculation of Neutron Currents in Spherical and Cylindrical Cavities by Means of View Factors. |
MUSPALB
|
Abstract
|
C00171 ICL00 00 |
Albedo Calculation of Multigroup Spectra of Neutrons Transmitted Through Multilayer Slab Shielding. |
MUXS |
Abstract
|
P00187 I3033 00 |
Generator of Multigroup Cross Sections for Charged Particle Transport Problems. |
MVP-GMVP II
|
Abstract
|
C00739 MNYCP 00 |
General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods. |
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. |
NAAPRO
|
Abstract
|
C00722 PC586 00 |
Neutron Activation Analysis PRognosis and Optimization Code System. |
NAB |
Abstract
|
D00018 I0360 00 |
100-Group, P3, Neutron Cross Section Data for Sodium and Aluminum. |
NAC
|
Abstract
|
C00164 C0000 00 |
Neutron Activation Analysis and Product Isotope Inventory Code System. |
NAC
|
Abstract
|
C00164 IBMMF 00 |
Neutron Activation Analysis and Product Isotope Inventory Code System. |
NAC-PC
|
Abstract
|
C00164 IBMPC 00 |
Neutron Activation Analysis and Product Isotope Inventory Code System. |
NACT
|
Abstract
|
C00502 U1100 00 |
Screening Program for Neutron Activation Products. |
NAISAP |
Abstract
|
P00085 F2306 00 |
Theory and Use of Gamma-Ray Spectrum Analysis Codes for NaI(Tl) Detectors. |
NANICK |
Abstract
|
P00120 I0360 00 |
Infinitely-Diluted Multigroup Cross-Section Generator - from ENDF/B. |
NAP
|
Abstract
|
C00101 I7090 00 |
Multigroup Time-Dependent Neutron Activation Prediction Code. |
NASIF-NARES |
Abstract
|
P00121 I0360 00 |
A Code System for Computing Shielding Factors from ENDF/B Tapes. |
NCRP49
|
Abstract
|
C00462 IBMPC 00 |
X-Ray Shield Calculation System. |
NCSP-DAT
| Abstract
|
M00002 MNYCP 01 |
Nuclear Data in Support of the Nuclear Criticality Safety Program. |
NE-SPEC |
Abstract
|
P00150 F2307 00 |
A Code System for Unfolding a Pulse Height Distribution of Neutrons Measured by an NE-213 Organic Scintillator. |
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 |
NEUPAC |
Abstract
|
P00177 FM200 00 |
Neutron Unfolding Code System for Calculating Neutron Flux Spectra from Activation Data of Dosimeter Foils. |
NEVEMOR |
Abstract
|
P00026 I3675 00 |
Multigroup-Multiregion Calculation of Flux Spectra and Energy Deposition for Fast Neutrons. |
NITRAN
|
Abstract
|
C00582 FM380 00 |
Neutron Transport Code System Based On Anisotropic Scattering. |
NJOY-UTIL-EIR |
Abstract
|
P00296 C0825 00 |
Utilities For the NJOY (6/83) Nuclear Data Processing System. |
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. |
NMTC/JAERI97
|
Abstract
|
C00694 SUN05 00 |
Monte Carlo Nucleon Meson Transport Code System. |
NMTC/JAM
|
Abstract
|
C00717 PC586 00 |
High Energy Particle Transport Code System. |
NONSAP-C |
Abstract
|
P00458 C7600 00 |
Code System for Analysis of 3-D Reinforced Concrete Structures. |
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. |
NOX |
Abstract
|
D00017 I0360 00 |
199-Group, P5, Coupled Neutron and Secondary Gamma-Ray Cross Section Data for Nitrogen and Oxygen. |
NPCSL-81 |
Abstract
|
D00082 I0370 00 |
Point Neutron Cross Sections Generated from ENDF/B-IV with the NPTXS Modules of PSR-63/AMPX-II. |
NPTXS |
Abstract
|
P00090 I0360 00 |
Data Generator: Neutron Point Cross Sections from ENDF/B Resolved and Unresolved Resonance Parameters. |
NRCDOSE 2.3.13
|
Abstract
|
C00684 PC586 10 |
Code System for Evaluating Routine Radioactive Effluents from Nuclear Power Plants with a Windows Interface. |
NRCPAGE |
Abstract
|
P00491 DVX11 00 |
Code System to Detect Recurring Loss of Special Nuclear Materials. |
NRCPIPES 2.0A |
Abstract
|
P00429 IBMPC 00 |
Code System for Fracture Mechanics Analysis of Circumferential Surface Cracks in Pipes. |
NRN
|
Abstract
|
C00054 C6600 00 |
Multigroup Removal-Diffusion Code System for Planes, Cylinders and Spheres. |
NSLINK |
Abstract
|
P00314 D0VAX 00 |
NJOY SCALE LINK. |
NUCCON
|
Abstract
|
C00439 S7800 00 |
A Code System for Calculation of Time-Dependent Nuclide Concentrations, Activity, Gamma-Ray Dose Rate and Biological Hazard Potential of Fusion Reactor Materials Due to Neutron Irradiation. |
NUCDECAY |
Abstract
|
D00172 PC386 01 |
Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP and MIRD. |
NUCDECAYCALC |
Abstract
|
D00202 PC586 00 |
Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP. See newer version in RASCAL (CCC-553). |
NUCHART |
Abstract
|
P00545 IBMPC 00 |
Nuclear Properties and Decay Data Chart of Nuclides. |
NUFACE |
Abstract
|
P00284 CYXMP 00 |
An Interface Code For The Calculation of Nuclear Responses. |
NUGAM 2&3 SSLAB
|
Abstract
|
C00210 I0360 00 |
Monte Carlo Prediction of Photon Transport Distributions. |
NUTRAN
|
Abstract
|
C00675 I0370 00 |
Code System for Long-Term Repository Safety Analysis. |
NX1-NX2 |
Abstract
|
P00310 D0VAX 00 |
Code System to Calculate Excitation Functions for (n,charged particle) Reactions. |
O5R
|
Abstract
|
C00017 I3675 00 |
A General-Purpose Monte Carlo Neutron Transport Code System. |
O5S |
Abstract
|
P00014 DP010 00 |
Response Function Generator--An O5R Monte Carlo Code for Calculating Pulse Height Distributions Due to Monoenergetic Neutrons Incident on Organic Scintillators. |
O5S |
Abstract
|
P00014 I3675 00 |
Response Function Generator--An O5R Monte Carlo Code for Calculating Pulse Height Distributions Due to Monoenergetic Neutrons Incident on Organic Scintillators. |
O6R
|
Abstract
|
C00128 I3675 00 |
A General-Purpose Monte Carlo Transport Code System. |
OCA-P |
Abstract
|
P00392 I3033 00 |
Pressure Vessel Fracture-Mechanics Code System. |
OCA-P |
Abstract
|
P00392 IBMPC 00 |
Pressure Vessel Fracture-Mechanics Code System. |
OCTAVIA |
Abstract
|
P00460 I0370 00 |
Code System to Calculate Pressure Vessel Failure Probabilities. |
OGRE
|
Abstract
|
C00046 I3675 00 |
A General-Purpose Monte Carlo Gamma-Ray Transport Code System. |
OGRE-MIN
|
Abstract
|
C00409 DGECL 00 |
A General-Purpose Monte Carlo Gamma-Ray Transport Code System for Minicomputers. |
OMCOST |
Abstract
|
P00381 I3033 00 |
Code System for Non-fuel O & M Cost Estimation for Large Steam-Electric Power Plants. |
OMEGA
|
Abstract
|
C00433 BESM6 00 |
Monte Carlo Criticality Code System. |
ONETRAN
|
Abstract
|
C00266 C7600 00 |
A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System. We recommend CCC-547/TWODANT-SYS. |
ONETRAN
|
Abstract
|
C00266 CY000 00 |
A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System. We recommend CCC-547/TWODANT-SYS. |
ONETRAN
|
Abstract
|
C00266 I3033 00 |
A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System. We recommend CCC-547/TWODANT-SYS. |
OOSII
|
Abstract
|
C00324 C0000 00 |
Calculation of Isotropic Scattering by Particles for One-Dimensional and Three-Dimensional Transport in Slabs by Invariant Imbedding, Orders-of-Scattering Method, Including Check Calculations by Integral Transport Theory and Monte Carlo. |
OPEX-II
|
Abstract
|
C00103 I7090 00 |
Radiation Shield Optimization Code. |
ORCENT-2 |
Abstract
|
P00474 I3033 00 |
Code System for Analysis of Steam Turbine Cycles Supplied by Light Water Reactors. |
ORIGEN-JENDL32
|
Abstract
|
C00703 MNYWS 00 |
Isotope Generation and Depletion Code with Libraries Based on JENDL3.2. New ORIGEN users are advised to get CCC-750/SCALE6 and run the ORIGEN-ARP code system in that package. |
ORIGEN2.2
|
Abstract
|
C00371 ALLCP 03 |
Isotope Generation and Depletion Code - Matrix Exponential Method. New ORIGEN users are advised to get CCC-750/SCALE6 and run the ORIGEN-ARP code system in that package. |
ORINC
USSO
|
Abstract
|
P00439 I0360 00 |
Code System for 1-D Implicit Heat Conduction Solution. |
ORION-II
|
Abstract
|
C00491 FM780 00 |
A Computer Code to Estimate Environmental Concentration and Dose Due to Airborne Release of Radioactive Material. |
ORIP_XXI
|
Abstract
|
C00731 PC586 01 |
Computer Programs for Isotope Transmutation Simulations. |
ORMDIN
USSO
|
Abstract
|
P00399 I3033 00 |
2-D Nonlinear Inverse Heat Conduction. |
ORMGEN3D |
Abstract
|
P00430 CY0MP 00 |
Mesh Generator for 3-D Crack Geometries. |
ORMONTE |
Abstract
|
P00275 IBMPC 00 |
Uncertainty Analysis Code System for Use with User-Developed Systems Models. |
ORPHEE VI
|
Abstract
|
C00159 I3675 00 |
Kernel Integration Code System - Attenuation of Fast Neutrons in Cylindrical Layers of Water and Dense Material. |
ORPLOT-PC |
Abstract
|
P00328 PC386 00 |
Plotting Package for Data Evaluation Intercomparison. |
ORSMAC
USSO
|
Abstract
|
P00437 I3033 00 |
Code System to Calculate Fluid Circulation Patterns Near Jets. |
ORTURB |
Abstract
|
P00418 I0360 00 |
HTGR Steam Turbine Dynamic Behavior. |
ORYX-E |
Abstract
|
D00038 I0360 00 |
ORIGEN Yields and Cross Sections Nuclear Transmutation and Decay Data from ENDF/B-IV. |
ORYX-E |
Abstract
|
D00038 I0360 01 |
ORIGEN Yields and Cross Sections Nuclear Transmutation and Decay Data from ENDF/B-IV. |
OZMA
|
Abstract
|
C00406 I0370 00 |
Calculation of Resonance Reaction Rates in Reactor Lattices Using Resonance Profile Tabulations. |
P-CARES |
Abstract
|
P00538 PC586 00 |
Probabilistic Computer Analysis for Rapid Evaluation of Structures. |
PABLM
|
Abstract
|
C00402 U1100 00 |
Calculation of Accumulated Radiation Doses to Man from Radionuclides Found in Food Products and from Radionuclides in the Environment. |
PADLOC
|
Abstract
|
C00330 U0000 00 |
A One-Dimensional, Time-Dependent Program for Calculating Coolant and Plateout Fission Product Concentrations in a Network of Pipes. |
PAGAN
|
Abstract
|
C00621 IBMPC 00 |
Code System for Performance Assessment Ground-water Analysis for Low-level Nuclear Waste. |
PALLAS-1D(VII)
|
Abstract
|
C00380 FM380 00 |
Multigroup Time-Independent Neutron Transport Code System for Plane or Spherical Geometry. |
PALLAS-2DCY-FX
|
Abstract
|
C00391 FM380 00 |
Multigroup Neutron/Gamma-Ray Direct Integration Transport Code System for Two-Dimensional Cylindrical Geometry. |
PAPER 1 |
Abstract
|
P00097 C6600 00 |
Monte Carlo Calculation of Solid Angle and Self-Absorption Factors for an Inclined Cylindrical Source Viewed by a Cylindrical Detector. |
PAPIN |
Abstract
|
P00156 I0370 00 |
A Code System to Calculate Cross Section Probability Tables, Bondarenko and Transmission Self-Shielding Factors for Fertile Isotopes in the Unresolved Resonance Region. |
PARET-ANL |
Abstract
|
P00516 MNYCP 00 |
Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores. |
PART61
|
Abstract
|
C00499 IBMPC 01 |
Low-Level Radioactive Waste Impacts Analysis System. |
PARTISN 4.0
|
Abstract
|
C00707 MNYCP 01 |
Time-Dependent, Parallel Neutral Particle Transport Code System. |
PATCH-7
|
Abstract
|
C00243 C0074 00 |
Three-Dimensional Kernel Integration Code-Explicit Single Scattering Option. |
PAVAN
|
Abstract
|
C00445 I3033 00 |
Atmospheric Dispersion Code System for Evaluating Accidental Radioactivity Releases from Nuclear Power Stations. |
PC-BATLE |
Abstract
|
P00451 IBMPC 00 |
Code System to Calculate Brief Adversary Threat Loss Estimate. |
PC-PRAISE |
Abstract
|
P00391 IBMPC 00 |
Code System for Analysis of Piping Reliability Including Seismic Events. |
PCC/SRC |
Abstract
|
P00456 D0VAX 00 |
Code System to Calculate Correlation & Regression Coefficients. |
PCDOSE
FEDC
|
Abstract
|
C00630 IBMPC 00 |
Radioactive Dose Assessment and NRC Verification of Licensee Dose Calculation. |
PEFPYD |
Abstract
|
D00096 ALLMF 02 |
Aggregate Fission-Product Decay Data Based on ENDF/B-IV and -V. |
PEGAS |
Abstract
|
P00336 IBMPC 00 |
Pre-Equilibrium-Equilibrium Gamma-and-Spin Code System. |
PELE-1C |
Abstract
|
P00461 C7600 00 |
Code System for Fluid-Structure Interaction Analysis. |
PELINSCA |
Abstract
|
P00168 I0360 00 |
A Code System for Nuclear Elastic and Inelastic Scattering Calculations. |
PELSHIE
|
Abstract
|
C00202 C0000 00 |
General Purpose Kernel Integration Shielding Code System-Point and Extended Gamma-Ray Sources. |
PELSHIE3
|
Abstract
|
C00202 IBMMF 00 |
General Purpose Kernel Integration Shielding Code System-Point and Extended Gamma-Ray Sources. |
PENELOPE-MPI
|
Abstract
|
C00713 IBMSP 00 |
Code System to Perform Monte Carlo Simulation of Electron Gamma-Ray Showers in Arbitrary Marerials. |
PENELOPE2008
|
Abstract
|
C00756 PC586 00 |
Code System to Perform Monte Carlo Simulation of Electron Photon Showers in Arbitrary Marerials. |
PEPIN
|
Abstract
|
C00285 I0360 00 |
Methodology for Computing Concentrations, Activities, Gamma-Ray Spectra, and Residual Heat from Fission Products. |
PEQAG-2 |
Abstract
|
P00293 IPCAT 00 |
A Pre-equilibrium Computer Code With Gamma Emission. |
PF-COMP
|
Abstract
|
C00106 C3600 00 |
Building Fallout Radiation Protection Factor Analysis. |
PFPL
|
Abstract
|
C00607 D0VAX 00 |
Puff-Plume Atmospheric Deposition Model. |
PHAZE
USSO
|
Abstract
|
P00432 IBMPC 00 |
Parametric Hazard Function Estimation. |
PHOEL-2
|
Abstract
|
C00327 I0360 00 |
A Monte Carlo Calculation of Initial Energy of Photoelectrons and Compton Electrons Produced by Photons in Water. |
PHOTX |
Abstract
|
D00136 IBMPC 00 |
Photon Interaction Cross Section Library. |
PHOTX |
Abstract
|
D00136 D0VAX 01 |
Photon Interaction Cross Section Library. |
PICA
|
Abstract
|
C00160 D0VAX 00 |
Monte Carlo Medium-Energy Photon-Induced Intranuclear Cascade Anal Code System. |
PICA
|
Abstract
|
C00160 I0360 00 |
Monte Carlo Medium-Energy Photon-Induced Intranuclear Cascade Anal Code System. |
PICFEE
|
Abstract
|
C00175 I3675 00 |
Fission Product Inventory Code System. |
PICTURE |
Abstract
|
P00238 IBMPC 00 |
Combinatorial Geometry Printer Plotting. |
PIEDEC
|
Abstract
|
C00566 FM380 00 |
A Practical Internal Exposure Dose Evaluation Code. |
PIGG
|
Abstract
|
C00138 C3600 00 |
A Multigroup One-Dimensional P-1 Radiation Transport Code System. |
PIPE
|
Abstract
|
C00219 I0360 00 |
Numerical Gamma-Ray Transport Code System for Plane/Spherical Geometry. |
PIXSE |
Abstract
|
P00133 I0360 00 |
A Generator of Multigroup and Multipoint Cross Sections for Thermal Reactor Calculations. |
PKI
|
Abstract
|
C00573 C0830 00 |
A Point Kernel Integration Code For Radiation Shielding of Loop System. |
PLACID
|
Abstract
|
C00381 I0370 00 |
Monte Carlo Simulation of Gamma Streaming Through Straight Cylindrical Ducts. |
PLASMX |
Abstract
|
P00106 C6600 00 |
A Multigroup Ionization and Charge Exchange Cross-Section Code System for Neutral Hydrogen Transport in Plasmas. |
PLOTENDF |
Abstract
|
P00214 I3033 00 |
A Program for Producing Graphical Output. |
PLOTFB |
Abstract
|
P00018 I3675 00 |
ENDF/B Data Plotting Code. |
PLOTNFIT |
Abstract
|
P00382 IBMPC 00 |
Code System for Data Plotting and Curve Fitting. |
PLOTTAB-89.1 |
Abstract
|
P00274 ALLCP 00 |
Plot Continuous Curves or Discrete Points. |
PLUDOS
|
Abstract
|
C00313 I0360 00 |
Calculator of Ground Level External Gamma-Ray Dose from a Radioactive Plume. |
PLUMEX
|
Abstract
|
C00356 I0360 00 |
A Computer Program to Evaluate External Exposures to a Gaussian Plume by Point Kernel Integration. |
PNESD |
Abstract
|
D00166 PC386 00 |
Proton Nucleus Elastic Scattering Data. |
POINT2000 |
Abstract
|
D00212 MNYCP 00 |
A Temperature-Dependent Linearly Interpolable, Tabulated Cross Section Library Based on ENDF/B-VI, Release 7. |
POINT2003 |
Abstract
|
D00218 MNYCP 00 |
A Temperature-Dependent Linearly Interpolable, Tabulated Cross Section Library Based on ENDF/B-VI, Release 8. |
POINT2004 |
Abstract
|
D00219 MNYCP 00 |
A Temperature-Dependent Linearly Interpolable, Tabulated Cross Section Library Based on ENDF/B-VI, Release 8. |
POINT2009 |
Abstract
|
D00239 MNYCP 00 |
A Temperature-Dependent Linearly Interpolable, Tabulated Cross Section Library Based on ENDF/B-VII.0 |
POINT97 |
Abstract
|
D00192 MNYCP 00 |
A Temperature-Dependent, Linearly Interpolable, Tabulated Cross Section Library Based on ENDF/B-VI, Release 4. |
POLLA |
Abstract
|
P00208 I3033 00 |
A Fortran Program to Convert R-MATRIX-Type Multilevel Resonance Parameters for Fissile Nuclei into Equivalent KAPUR-PEIERLS-Type Parameters. |
POLYRES |
Abstract
|
P00438 MNYCP 00 |
Richards Equation Solver; Rectangular Finite Volume Flux Updating Solution. |
POPLIB |
Abstract
|
D00012 I0360 03 |
A Compendium of Neutron-Induced Secondary Gamma-Ray Yield and Cross Section Data. |
POPOP4 |
Abstract
|
P00011 I3675 00 |
Converter of Gamma-Ray Spectra to Secondary Gamma-Ray Production Cross Sections. |
POWER |
Abstract
|
P00069 C7600 00 |
Source Distribution Input Data Generator for ANISN Code. |
PR-EDB |
Abstract
|
D00196 IBMPC 03 |
Power Reactor Embrittlement Data Base, Version 3. |
PRAISE-C |
Abstract
|
P00391 C7600 00 |
Code System for Analysis of Piping Reliability Including Seismic Events. |
PRE-ANISN |
Abstract
|
P00332 PC386 00 |
A Preprocessing Code for ANISN and Other Radiation Transport Codes. |
PREANG |
Abstract
|
P00166 C0175 00 |
Calculation of Pre-equilibrium Angular Distributions with the Exciton Model. |
PRECO2006 |
Abstract
|
P00226 MNYCP 02 |
Exciton Model Code System for Calculating Preequilibrium and Direct Double Differential Cross Sections. |
PREM |
Abstract
|
P00224 I0360 00 |
Code System for Pre-equilibrium Process with Multiple Nucleon Emission. |
PREMOR
|
Abstract
|
C00369 I0360 00 |
A Point Reactor Exposure Code System for Survey Nuclear Analysis of Power Plant Performance. |
PREPRO2007 |
Abstract
|
P00351 MNYCP 05 |
Pre-processing Code System for Data in ENDF/B Format. |
PREST
|
Abstract
|
C00355 I0360 00 |
Calculator of Pressure and Temperature Transient in Containment Studies. |
PRESTO
|
Abstract
|
C00549 D8810 00 |
Point Kernel Calculation for Complex and Time-Dependent Gamma-Ray Source Spectra. |
PRESTO-II
|
Abstract
|
C00504 I0360 00 |
Code System for Low-Level Waste Environmental Transport and Risk Assessment. |
PRIMEDANA-2
|
Abstract
|
C00490 I3081 00 |
Collapses Multigroup Cross Sections and Obtains Reaction Parameters by Solving Transport or Diffusion Equations. |
PRISIM
|
Abstract
|
C00574 IBMPC 00 |
Plant Risk Status Information Management System. |
PROB
|
Abstract
|
C00287 I0370 00 |
Multigroup One-Dimensional Transport Code System, Collision Probability Method. |
PROCIV
|
Abstract
|
C00488 U1110 00 |
A Code System for Calculating the Protection Factors Against Radioactive Fallout for Apartment Buildings. |
PSDREC |
Abstract
|
P00441 DP011 00 |
Code System for Power Spectral Density Recognition Continuous On-line Reactor Surveillance. |
PSU-LEOPARD/RBI
|
Abstract
|
C00563 IBMPC 01 |
A Spectrum Dependent Non-Spatial Depletion Code. |
PTRAN
|
Abstract
|
C00618 PC386 00 |
Proton Monte Carlo Transport Program for the PC. |
PUCOR |
Abstract
|
D00067 I3691 00 |
84 Group Neutron Cross Sections for Uranium-Plutonium Cycle LWR and PWR Models in AMPX Master Library Format. |
PUDK |
Abstract
|
D00074 I0360 00 |
Measured Results of Delayed Beta- and Gamma-Ray Spectra Due to Thermal-Neutron Fission of Pu239 and Pu241. |
PUFF-IV |
Abstract
|
P00534 MNYCP 01 |
Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files, Version 6.0.1. |
PURSE
|
Abstract
|
C00338 C6600 00 |
A Plutonium Radiation Source Code System. |
PUSHLD
|
Abstract
|
C00271 C0074 00 |
Gamma-Ray Three-Dimensional Calculation of Dose Rates from Plutonium in Various Geometries. |
PUTZ 2.1
|
Abstract
|
C00595 IBMPC 00 |
A Point-Kernel Photon Shielding Code. |
PVC |
Abstract
|
D00048 I3691 00 |
36 Group, P5, Photon Interaction Cross Sections for 38 Materials in ANISN Format. |
PVE |
Abstract
|
D00126 I3033 00 |
38 Group, P8, Photon Interaction Cross Sections in ANISN Format from VITAMIN-E. |
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. |
Q&A |
Abstract
|
P00428 IBMPC 00 |
Questions and Answers Based on Revised 10 CFR Part 20 |
QAD
|
Abstract
|
C00048 I0360 00 |
Kernel Integration Code System. |
QAD-BSA
|
Abstract
|
C00346 C0000 00 |
Point-Kernel Shielding Code System. |
QAD-CGGP-A
|
Abstract
|
C00645 MNYCP 00 |
Point Kernel Code System for Neutron and Gamma-Ray Shielding Calculations Using the GP Buildup Factor. |
QAD-P5
|
Abstract
|
C00048 C6400 00 |
Kernel Integration Code System. |
QAD-QC
|
Abstract
|
C00401 C0000 00 |
Three-Dimensional Point Kernel Gamma-Ray Shielding Code. |
QAD-QC
|
Abstract
|
C00401 I0360 00 |
Three-Dimensional Point Kernel Gamma-Ray Shielding Code. |
QAD-UE
|
Abstract
|
C00448 H6000 00 |
A Revised Numerical Integration Option for Gamma-Ray Volume Source Problems in the QAD-CG Point Kernel Shielding Code. |
QADMOD-G
|
Abstract
|
C00396 I3033 00 |
Point Kernel Gamma-Ray Shielding Code. |
QADMOD-GP
|
Abstract
|
C00565 IBMPC 00 |
Point Kernel Gamma-Ray Shielding Code With Geometric Progression Buildup Factors. |
QBF
|
Abstract
|
C00617 PC386 00 |
Code System to Calculate Radiation Dose Rates Relative to Spent Fuel Shipping Casks. |
QBSHIELD
|
Abstract
|
C00599 IBMPC 00 |
Spherical Shield Design for Gamma-Ray Sources Using the Buildup Factor Method. |
QUARK |
Abstract
|
P00492 PC586 00 |
Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics. |
QUINCE-PC
|
Abstract
|
C00556 IBMPC 00 |
Calculates Absorbed Dose From Skin Contamination. |
RABFIN PARTS
|
Abstract
|
C00668 IBMPC 00 |
Code System for Calculating Gaseous Effluent Dose Parameters. |
RACC
|
Abstract
|
C00388 CY000 00 |
A Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems. |
RACC
|
Abstract
|
C00388 I3033 00 |
A Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems. |
RACC-PULSE
|
Abstract
|
C00639 MNYWS 00 |
RACC Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems Modified for Pulsed/Intermittent Activation Analysis. |
RACER
|
Abstract
|
C00174 U1108 00 |
Calculation of Potential External Dose from Airborne Fission Products Following Postulated Reactor Accident. |
RAD 2
|
Abstract
|
C00122 I7090 00 |
Fission Product Radioactivities Calculation. |
RADAC
|
Abstract
|
C00627 PC486 02 |
Code System for Calculating Radioactive Decay and Accumulation of Decayed Products Using Integer-Array Arithmetic for Precise Evaluation of the Bateman Equations. |
RADAK |
Abstract
|
P00122 I0360 00 |
Flux Spectra Unfolding Code System - Neutron or Gamma-Ray Detectors. |
RADCOMPT 2.10L |
Abstract
|
P00348 IBMPC 00 |
Sample Analysis Code System for the Dual Channel Counter. |
RADDECAY 4.02 |
Abstract
|
D00134 IBMPC 03 |
Radioactive Decay Data for Radiological Assessments. |
RADHEAT-V4
|
Abstract
|
C00300 FM380 00 |
A Code System To Generate Multigroup Constants and Analyze Radiation Transport for Shielding Safety Evaluation. |
RADOS
|
Abstract
|
C00088 I3675 00 |
Gamma-Ray Dose Estimation from Cloud of Radioactive Gases by Kernel Integration. |
RADRISK
|
Abstract
|
C00422 DGMV1 00 |
Estimates Radiation Doses and Health Effects from Inhalation or Ingestion of Radionuclides. See C00476/CAAC. |
RADRISK
|
Abstract
|
C00422 I3033 00 |
Estimates Radiation Doses and Health Effects from Inhalation or Ingestion of Radionuclides. See C00476/CAAC. |
RADSHIP-2
|
Abstract
|
C00523 FM200 00 |
Code System To Analyze Radiological Impact From Radwaste Transportation. |
RADSYS
|
Abstract
|
C00530 I3033 00 |
Code System for Radioactivity Buildup and Radioactive Waste Generation Calculations. |
RAFFLE/2
|
Abstract
|
C00279 C0176 00 |
A General Purpose Monte Carlo Code System for Neutron Transport with Mixed Zone Geometry Option. |
RAFFLE/2 MOD 2
|
Abstract
|
C00279 I0360 00 |
A General Purpose Monte Carlo Code System for Neutron Transport with Mixed Zone Geometry Option. |
RAID
|
Abstract
|
C00083 I7090 00 |
Monte Carlo Multibend Duct Shielding Code. |
RANCHMD
|
Abstract
|
C00589 D8810 00 |
Radionuclide Chain Transport with Matrix Diffusion. |
RASC-2D
|
Abstract
|
C00318 I0370 00 |
Two-Dimensional Removal Diffusion Code Reactor Shielding Design Code System. |
RASCAL 3.0.5
|
Abstract
|
C00553 PC586 10 |
Radiological Assessment for Consequence Analysis for Windows. |
RASPA
|
Abstract
|
C00352 C7600 00 |
A Code for the Calculation of Buildup and Decay of Fission Products and Actinides. |
RATAF
|
Abstract
|
C00681 IMFPC 01 |
Code System for the Radioactive Liquid Tank Failure Study. |
RBD
|
Abstract
|
C00632 IBMPC 00 |
U.S. Army Radiological Bioassay and Dosimetry. |
RCSLK9 |
Abstract
|
P00452 IBMPC 00 |
Code System to Calculate Reactor Coolant System Leak Rate. |
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. |
REACTION |
Abstract
|
P00347 AL000 00 |
Code System to Calculate Integral Parameters with Reaction Rates from WIMS Output. |
REACTION |
Abstract
|
P00347 IBMPC 00 |
Code System to Calculate Integral Parameters with Reaction Rates from WIMS Output. |
REBEL 3
|
Abstract
|
C00299 I0360 00 |
Adjoint Monte Carlo Calculation of Radiation Doses to Human Organs in Dwelling Rooms. |
REBEL-2
|
Abstract
|
C00299 C6600 00 |
Adjoint Monte Carlo Calculation of Radiation Doses to Human Organs in Dwelling Rooms. |
REBEL-2
|
Abstract
|
C00299 ICL00 00 |
Adjoint Monte Carlo Calculation of Radiation Doses to Human Organs in Dwelling Rooms. |
REBUS-PC 1.4
|
Abstract
|
C00708 PC586 00 |
Code System for Analysis of Research Reactor Fuel Cycles. |
REBUS3/VARIANT8
|
Abstract
|
C00653 MNYWS 01 |
Code System for Analysis of Fast Reactor Fuel Cycles. |
RECAP |
Abstract
|
P00414 IBMPC 00 |
Replacement Energy Cost Analysis Package. |
RECAP |
Abstract
|
P00414 IBMPC 01 |
Replacement Energy Cost Analysis Package. |
RECOIL |
Abstract
|
D00055 I3033 01 |
Multigroup Primary Recoil Spectra, Displacement Rates and Gas-Production Rates for Radiation Damage Studies. |
REDIFFUSION
|
Abstract
|
C00347 I0360 00 |
One-Dimensional Neutron Removal-Diffusion and Gamma-Ray Kernel Integration or Diffusion Theory Calculator. |
REFCO83 |
Abstract
|
P00447 I3033 00 |
Nuclear Fuel Cycle Cost Economics Code System. |
REFERDOU |
Abstract
|
P00249 FM380 00 |
Code System for NE-213 Unfolding of Neutron Spectra up to 100 MeV with Response Function Error Propagation. |
REFLUX |
Abstract
|
P00403 I3033 00 |
Code System to Predict LWR Reflood Heat Transfer. |
REFREP
|
Abstract
|
C00570 D8810 00 |
A Near-Field Model For A Spent Fuel Repository. |
REFUM-BROAD |
Abstract
|
P00039 F2307 00 |
Monte Carlo Codes for Calculating Efficiencies and Response Functions of NaI(Tl) Crystals for Thick Disk Gamma-Ray Sources. |
REGN |
Abstract
|
P00165 I0360 00 |
Code System for Solving Nonlinear Systems of Equations via the Gauss-Newton Method. |
RELAP3B/MOD110 |
Abstract
|
P00422 C7600 00 |
Reactor System Transient Code. |
RELAP4/MOD7/101 |
Abstract
|
P00416 C0176 00 |
Best Estimate Code System to Calculate Thermal & Hydraulic Phenomena in a Nuclear Reactor or Related System. |
RELAP5/MOD1/025
USSO
|
Abstract
|
P00423 DVX11 00 |
Thermal Hydraulic Computer Code System. |
RELAP5/MOD1/025
USSO
|
Abstract
|
P00423 I3033 00 |
Thermal Hydraulic Computer Code System. |
RELAP5/MOD1/029
USSO
|
Abstract
|
P00423 C0176 00 |
Thermal Hydraulic Computer Code System. |
REMIT 5.1 |
Abstract
|
P00482 IBMPC 01 |
Radiation Exposure Monitoring and Information Transmittal System. |
REPC |
Abstract
|
P00195 C0000 00 |
Estimation of Nuclear Reaction Effects in Proton-Tissue-Dose Calculations. |
REPRISK PC 1.02
|
Abstract
|
C00586 PC386 01 |
Repository Risk Assessment Software for Personal Computers. |
RESENDD |
Abstract
|
P00215 C0740 00 |
A Code System for Reconstruction of Resonance Cross Sections from Evaluated Nuclear Data in ENDF/B Format. |
RESENDD |
Abstract
|
P00215 D0780 00 |
A Code System for Reconstruction of Resonance Cross Sections from Evaluated Nuclear Data in ENDF/B Format. |
RESPMG |
Abstract
|
P00060 I0360 00 |
Response Matrix Generation Code System. |
RESRAD 5.82
|
Abstract
|
C00552 PC386 05 |
Code System to Implement Residual Radioactive Material Guidelines. |
REST 1;2;3
|
Abstract
|
C00225 I0360 00 |
Fission Product Inventory Code System with Fission Product Escape Model. |
RETRAC
|
Abstract
|
C00635 D0VAX 00 |
Code System for the Analysis of Material Test Reactor (MTR) Cores. |
RETRANS
|
Abstract
|
C00669 SUN05 00 |
Code System For Calculating Reactivity Transients In a LWR. |
REX2-87 |
Abstract
|
P00290 D8810 00 |
A Code For Calculating Self-Shielded Multigroup Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data Files. |
RFSP-JUL |
Abstract
|
P00126 I0360 00 |
Unfolding Code System for Neutron Spectra Evaluation from Activation Data. |
RFUNC |
Abstract
|
P00312 D0VAX 00 |
Code System to Analyze Differential Scattering Data. |
RGENDF |
Abstract
|
P00239 C0170 00 |
Format Translation from NJOY GENDF Format to ENDF/B-V and Other Formats. |
RHEIN
|
Abstract
|
C00585 I3090 00 |
Reactor Code System for Neutron Physics Calculation. |
RIBD-II
|
Abstract
|
C00137 C6600 00 |
Radioisotope Buildup and Decay Code System. |
RIBD-II
|
Abstract
|
C00137 I0360 00 |
Radioisotope Buildup and Decay Code System. |
RIBD-IRT
|
Abstract
|
C00382 U1100 00 |
Radioisotope Buildup and Decay Code System. |
RICANT
|
Abstract
|
C00569 D8810 00 |
A Computer Code for 2-D Transport Calculations in x-y Geometry Using the Interface Current Method. |
RICE |
Abstract
|
P00022 I0360 00 |
A Program to Calculate Primary Recoil Atom Spectra from ENDF/B Data. |
RICECCC
|
Abstract
|
C00348 I0360 00 |
A Reactor Nuclide Inventory Code for Calculating Actinides and Fission Products. |
RISKAP
|
Abstract
|
C00486 I3033 00 |
Analysis of Increased Risk to Arbitrary Populations. |
RISKIND 2.0
FEDC
|
Abstract
|
C00623 IBMPC 02 |
Radiological Risk Assessment Code System for Spent Nuclear Fuel Transportation. |
RITTS |
Abstract
|
D00011 I0360 00 |
121-Group Coupled Neutron and Gamma-Ray Cross-Section Data for Transport Codes. |
RIVER-RAD
|
Abstract
|
C00626 MNYCP 00 |
Code System for Simulating the Transport of Radionuclides in Rivers. |
RMET21
|
Abstract
|
C00597 D0VAX 00 |
Detailed Space and Energy Treatment of Neutron Resonances for Homogeneous Mixtures and Cylinderized Reactor Cells. |
RNGP |
Abstract
|
P00066 I3675 00 |
Random Number Generator Package. |
ROLAIDS-CPM |
Abstract
|
P00353 SUN04 00 |
Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method. |
RRR
|
Abstract
|
C00196 I0360 00 |
Radiation Transport in Air-Analysis of Routine Releases of Short-Lived Radioactive Nuclides. |
RSAC-6
|
Abstract
|
C00125 PC386 02 |
Radiological Safety Analysis Code System. |
RSYST
|
Abstract
|
C00269 I0360 00 |
Integrated Modular Code System for Shielding and Reactor Physics Calculations. |
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. |
SABRINA 3.54 |
Abstract
|
P00242 MFMWS 02 |
Three-Dimensional Geometry Visualization Code System. |
SACHET
|
Abstract
|
C00571 D8810 00 |
A Computer Program To Evaluate The Dynamic Fission Product Inventories in the Multiple Compartment System of PWR's. |
SAFE-D/SAFE-R |
Abstract
|
P00496 MNYCP 00 |
Code System for the Analysis of Component Failure Data with a Compound Statistical Model. |
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 I3033 00 |
Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96. |
SAILOR |
Abstract
|
D00076 PC386 01 |
Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96. |
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 |
Abstract
|
P00158 MNYCP 12 |
Code System for Multilevel R-Matrix Fits to Neutron and Charged-Particle Cross-Section Data Using Bayes' Equations. |
SAMPO-LRC |
Abstract
|
P00186 C6600 00 |
Gamma-Ray Spectrum Analysis Code. |
SAMPO80 |
Abstract
|
P00204 DGNOV 00 |
Gamma-Ray Spectrum Analysis Method for Minicomputers. |
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. We recommend PSR-345/SNL-SAND-II. |
SAND-II-SNL |
Abstract
|
P00345 SUN04 00 |
Neutron Flux Spectra Determination by Multiple Foil Activation - Iterative 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. |
SARA 4.16
USSO
|
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
|
Abstract
|
C00750 MNYCP 00 |
Modular Code System for Performing Criticality and Shielding Analyses for Licensing Evaluation with ORIGEN-ARP (Source & Executables). |
SCALE 6-EXE
|
Abstract
|
C00750 MNYCP 01 |
Modular Code System for Performing Criticality and Shielding Analyses for Licensing Evaluation with ORIGEN-ARP (Executables - No Source). |
SCAMPI |
Abstract
|
P00352 MNYWS 00 |
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-2B |
Abstract
|
P00294 MNYCP 02 |
Code System for Calculating Total and Elastic Scattering Cross Sections Based on an Optical Model of the Spherical Nucleus, Versions SCAT-2 and SCAT-2B. |
SCINFUL |
Abstract
|
P00267 CY0MP 00 |
Scintillator Full Response to Neutron Detection. |
SCINFUL |
Abstract
|
P00267 D8600 00 |
Scintillator Full Response to Neutron Detection. |
SCOPE |
Abstract
|
P00210 I3033 00 |
Computer Code System for Shipping Cask Optimization and Parametric Evaluation. |
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. |
SEECAL 2.0
|
Abstract
|
C00620 IBMPC 00 |
Program to Calculate Age-Dependent Specific Effective Energies. |
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-1C0
|
Abstract
|
C00729 MNYCP 00 |
Simulation Environment for Radiotherapy Applications. |
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 SUN05 00 |
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 2009.02 |
Abstract
|
D00237 MNYCP 00 |
Shielding Integral Benchmark Archive and Database, Version February 2009. |
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. |
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 |
Code System to Calculate Neutron and Gamma-Ray Skyshine Doses Using the Integral Line-Beam Method. |
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. |
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. |
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 |
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. |
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 1.1
USSO
|
Abstract
|
P00533 PC586 00 |
Code System to Predict Leakage Rate and Area of Crack Opening for Cracked Pipes in Nuclear Power Plants, Version 1.1 for Windows. |
SRAC95
|
Abstract
|
C00716 MNYWS 00 |
Thermal Reactor Code System for Reactor Design and Analysis. |
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. |
STAPRE-H95 |
Abstract
|
P00325 MNYCP 01 |
Code System to Calculate Energy-Averaged Cross Sections of Particle Induced Nuclear Reactions. |
STAPREF |
Abstract
|
P00498 PC586 00 |
Code System to Calculate Nuclear Reaction Cross Sections by Evaporation Model. |
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. |
STERNO
|
Abstract
|
C00057 C0000 00 |
Two Dimensional Gamma-Ray Heating Kernel Integration Code. |
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. |
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 |
A Code System for Generating Transport Group Constants, Energy Deposition Coefficients and Atomic Displacement Constants with ENDF/B. |
SUPERTOG-JR. |
Abstract
|
P00115 I0360 00 |
A Code System for Generating Transport Group Constants, Energy Deposition Coefficients and Atomic Displacement Constants with ENDF/B. |
SUPERTOG-LTT |
Abstract
|
P00228 I0360 00 |
A Modification of PSR-13/SUPERTOG-III Applied to Libraries with Tabulated Elastic Scattering and Anistropy Densities. |
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. |
SWAT
|
Abstract
|
C00714 MNYWS 00 |
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 andTransport. |
SYVAC-D/2
|
Abstract
|
C00690 D0VAX 00 |
Code System For Risk Assessment From Underground Radioactive Waste Disposal In the United Kingdom. |
TACT-III
|
Abstract
|
C00447 I3033 00 |
Calculation of the Transport of Radioactivity from a Reactor Core. |
TALYS 1.0 |
Abstract
|
P00548 PC586 00 |
Nuclear Model Code System for Analysis and Prediction of Nuclear Reactions and Generation of Nuclear Data. |
TAM3 |
Abstract
|
P00308 IBMPC 00 |
Demonstrates Monte Carlo Sensitivity and Uncertainty Analysis. |
TART2005
|
Abstract
|
C00638 MNYCP 06 |
Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code System. |
TASK
|
Abstract
|
C00184 I0360 00 |
Generalized One-Dimensional Radiation Transport and Diffusion Kinetics Code System. |
TDA
|
Abstract
|
C00180 MNYWS 01 |
A Time-Dependent, Multigroup, One-Dimensional, Discrete Ordinates Transport Code System. |
TDF |
Abstract
|
D00162 ALLCP 00 |
Thermonuclear Data File. |
TDOWN-IV |
Abstract
|
P00172 H6000 00 |
A Code System to Generate Composition- and Spatially-Dependent Neutron Cross Sections for Multigroup Neutronics Analysis. |
TDT
|
Abstract
|
C00256 I0360 00 |
Generalized One-Dimensional Multigroup Time-Dependent Transport and Diffusion Kinetic Code System. |
TDTORT
|
Abstract
|
C00709 MNYWS 00 |
Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System. |
TECALC |
Abstract
|
P00074 DP010 00 |
Interactive Calculation of Compton Coherent and Photoelectric Mass Attenuation Coefficients for Photons (E<1 MeV), and the Mass Absorption Coefficient for Known Materials. |
TEMAC |
Abstract
|
P00468 D0VAX 00 |
Top Event Matrix Analysis Code System. |
TERFOC-N
|
Abstract
|
C00596 MFMWS 00 |
Terrestrial Food-Chain Model for Normal Operations. |
TESS
|
Abstract
|
C00215 C3600 00 |
Multigroup Discrete Ordinates Code System for Slab and Spherical Geometries. |
THERMGAM |
Abstract
|
D00140 ALLCP 00 |
Prompt Gamma Rays from Thermal-Neutron Capture. |
THERMOS-OTA |
Abstract
|
P00107 C0173 00 |
Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
THERMOS-OTA |
Abstract
|
P00107 C0740 00 |
Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
THERMOS-OTA |
Abstract
|
P00107 U1108 00 |
Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
THIDA-2
|
Abstract
|
C00410 FM380 00 |
Code System for the Calculation of Transmutation, Activation, Decay Heat and Dose Rate in Fusion Reactors. |
THRUSH |
Abstract
|
P00276 CYXMP 00 |
Calculates Thermal Neutron Scattering Kernel. |
THT
|
Abstract
|
C00480 I0360 00 |
Three-Dimensional Neutron Coarse Mesh Code System to Evaluate Average Bundle Fluxes and Power in Light Water Reactors. |
TIBSO
|
Abstract
|
C00512 MNYCP 00 |
Code System to Calculate Production and Migration of Radionuclides in Nuclear Reactor Systems. |
TIMED
|
Abstract
|
C00292 I0360 00 |
Calculation of Cumulated Activity of a Radionuclide in the Organs of the Human Body at a Given Time After Deposition. |
TIMEX
|
Abstract
|
C00274 C7600 00 |
One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering. |
TIMEX
|
Abstract
|
C00274 CY000 00 |
One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering. |
TIMEX
|
Abstract
|
C00274 U1106 00 |
One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering. |
TIMOC-72
|
Abstract
|
C00144 I0370 00 |
Monte Carlo Three-Dimensional Neutron Transport Code System. |
TIMOC-ESP
|
Abstract
|
C00432 U1110 00 |
System for Generating and Analyzing Time Dependent Radiation Transport Results by Monte Carlo. |
TIMS-1 |
Abstract
|
P00163 D0780 00 |
Processing Code System for Production of Group Constants of Heavy Resonant Nuclei. |
TIMS-1 |
Abstract
|
P00163 FM200 00 |
Processing Code System for Production of Group Constants of Heavy Resonant Nuclei. |
TIRION 4
|
Abstract
|
C00395 I3033 00 |
A Program for Calculating Consequences of a Release of Radioactive Material to the Atmosphere. |
TMMS
|
Abstract
|
C00246 I0360 00 |
Gamma-Ray Penetration Shielding Code System, Transmission Matrix Method. |
TNG1 |
Abstract
|
P00298 D6220 00 |
A Multistep Statistical Model Based on the Hauser-Feshbach Theory For The Evaluation Of Neutron Data. |
TORAC |
Abstract
|
P00459 C0170 00 |
Code System to Calculate Tornado-Induced Flow Material Transport. |
TOXRISK
|
Abstract
|
C00692 CDCMF 00 |
Code System for Toxic Gas Accident Analysis. |
TP1
|
Abstract
|
C00465 I3033 00 |
A Computer Code System for the Calculation of Reactivity and Kinetic Parameters by One-Dimensional Neutron Transport Perturbation Theory. |
TP2
|
Abstract
|
C00470 I3033 00 |
A Computer Program for the Calculation of Reactivity and Kinetic Parameters by Two-Dimensional Neutron Transport Perturbation Theory. |
TPASGAM 85 |
Abstract
|
D00088 ALLCP 04 |
Radioactive Decay Library of Gamma-Ray Energies, Branching Ratios, and Cross Sections. |
TPASS |
Abstract
|
P00164 DP010 00 |
A Gamma-Ray Spectral Data-Reduction and Analysis Code System. |
TPHEX
|
Abstract
|
C00421 C0173 00 |
Transmission Probability Code System for Calculating Neutron Flux Distributions in Hexagonal Geometry. |
TPHEX
|
Abstract
|
C00421 CYXMP 00 |
Transmission Probability Code System for Calculating Neutron Flux Distributions in Hexagonal Geometry. |
TPTRIA
|
Abstract
|
C00550 I3083 00 |
A Computer Program for the Reactivity and Kinetic Parameters for Two-Dimensional Triangular Geometry by Transport Perturbation Theory. |
TR-EDB |
Abstract
|
D00198 IBMPC 00 |
Test Reactor Embrittlement Data Base, Version 1. |
TRAC-BD1
USSO
|
Abstract
|
P00488 C0176 00 |
Code System for Best-Estimate Analysis of LOCA in BWR. |
TRAC-PF1
USSO
|
Abstract
|
P00481 IBMPC 00 |
Best-Estimate Analysis PWR LOCA. |
TRAC-PF1/EN MOD3 |
Abstract
|
P00477 PC486 00 |
Code System for Coupled 3D Neutronics-Thermalhydraulics Calculations. |
TRANSHEX
|
Abstract
|
C00449 U1108 00 |
Two-dimensional Multigroup Collision Probability Code System for Hexagonal Geometry. |
TRANSMIT |
Abstract
|
D00020 I0360 00 |
Experimental Neutron Transmission Data Used to Test Total Cross Sections. |
TRANSPORT
|
Abstract
|
C00244 C6600 00 |
Charged Particle Beam Transport Systems Design Code System (First- and Second-Order Matrix Multiplication). |
TRANSPORT
|
Abstract
|
C00244 I0360 00 |
Charged Particle Beam Transport Systems Design Code System (First- and Second-Order Matrix Multiplication). |
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. |
TRANSX-CTR |
Abstract
|
P00206 CY000 00 |
Interfaces MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis. |
TRANZIT
|
Abstract
|
C00172 C7600 00 |
Multigroup Time-Dependent Discrete Ordinates Radiation Transport Code System in (rho,z) Cylindrical Geometry. |
TRAPP
|
Abstract
|
C00205 I3691 00 |
Transport of Alpha Particles and Protons with all Nuclear Reaction Products Neglected. |
TRAX |
Abstract
|
P00280 C0720 00 |
A Program For Optics of Curved Crystal Neutron Spectrometers. |
TRD-3
|
Abstract
|
C00362 I3033 00 |
Two-Dimensional Removal-Diffusion Neutron Shielding Code System. |
TRECO
|
Abstract
|
C00116 I3675 00 |
An Orbital Integration Estimation of Trapped Radiation. |
TREEDE
|
Abstract
|
C00326 C0000 00 |
Monte Carlo Neutron Transport Code System Based on the Track Rotation Estimator. |
TRG-SGD
|
Abstract
|
C00025 C0000 00 |
Calculation of Secondary Gamma-Ray Dose Rate from a Nuclear Weapon Detonation-Monte Carlo Method. |
TRIDENT
|
Abstract
|
C00293 C7600 00 |
Two-Dimensional Multigroup Discrete Ordinates Transport Code System-(x,y) and (r,z) Geometries. |
TRIDENT
|
Abstract
|
C00293 I0360 00 |
Two-Dimensional Multigroup Discrete Ordinates Transport Code System-(x,y) and (r,z) Geometries. |
TRIDENT-CTR
|
Abstract
|
C00377 C0000 00 |
Two-Dimensional x-y and r-z Geometry Multigroup Transport Code System for Large Toroidal Reactors. |
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. |
TRIGON
|
Abstract
|
C00290 U1108 00 |
Two-Dimensional Multigroup Diffusion Code System-Trigonal or Hexagonal Mesh. |
TRIPLET
|
Abstract
|
C00230 C6600 00 |
Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System. |
TRIPLET
|
Abstract
|
C00230 C7600 00 |
Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System. |
TRIPLET
|
Abstract
|
C00230 I0360 00 |
Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System. |
TRIPOLI 4.4
OECD
|
Abstract
|
C00747 MNYWS 00 |
Code System for Coupled Neutron, Photon, Electron, Positron, 3-D, Time Dependent, Monte-Carlo, Transport Calculations. |
TRIPOS
|
Abstract
|
C00537 CY00I 00 |
Monte Carlo Ion Transport Analysis Code. |
TRISTAN
|
Abstract
|
C00511 HM280 00 |
Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System. |
TRISTAN-IJS |
Abstract
|
P00537 IBMPC 00 |
Steady-State Axial Temperature and Flow Velocity in Triga Channel. |
TRITAC
|
Abstract
|
C00560 D8810 00 |
A Three-Dimensional Transport Code For Eigenvalue Problems Using The Diffusion Synthetic Acceleration Method. |
TRUMP |
Abstract
|
P00522 MNYCP 01 |
Code System for Transient and Steady-State Temperature Distribution in Multidimensional Systems. |
TSORT |
Abstract
|
P00486 IBMPC 00 |
Automated Technique for Nuclear Plant Training Task Assignment. |
TWOTRAN
|
Abstract
|
C00195 C6600 00 |
Two-Dimensional Discrete Ordinates. We recommend CCC-547/TWODANT-SYS. |
TWOTRAN II
|
Abstract
|
C00222 C7600 00 |
Two-Dimensional Multigroup Discrete Ordinates Transport C System in (x,y), (r,theta), and (r,z) Geometries. |
TWOTRAN II
|
Abstract
|
C00222 I3691 00 |
Two-Dimensional Multigroup Discrete Ordinates Transport C System in (x,y), (r,theta), and (r,z) Geometries. |
TWOTRAN-SPHERE
|
Abstract
|
C00129 C6600 00 |
Multigroup Two-Dimensional Discrete Ordinates Transport Code System in Spherical Geometry. |
UDAD IX
|
Abstract
|
C00685 I0370 00 |
Uranium Dispersion & Dosimetry Model. |
UHS |
Abstract
|
P00390 IPS70 00 |
Ultimate Heat Sink Cooling Pond and Spray Pond Analysis Models. |
UKCTRI-81 |
Abstract
|
D00064 I0370 01 |
46-Group Neutron Cross Sections and Kerma Factors for Fusion Reactor Calculations. |
UKE-III |
Abstract
|
P00015 I3691 00 |
Cross Section Format Translator - UKNDL to ENDF/B. |
UKFY2 |
Abstract
|
D00171 IBMPC 00 |
UK Fission Product Yield Library, Version 2. |
UKNDL |
Abstract
|
D00039 I0370 00 |
United Kingdom Evaluated Neutron Cross-Section Data Library. |
UKNDL-81 |
Abstract
|
D00107 I3033 00 |
The Aldermaston Nuclear Data Library. |
UMG 3.3 |
Abstract
|
P00529 PC586 00 |
Unfolding with Maxed and Gravel. |
UMIBIO
|
Abstract
|
C00680 I3033 00 |
Code System to Model Uranium Mills Bioassay Dosimetry. |
UNF |
Abstract
|
P00521 PC586 00 |
Code System to Calculate Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials. |
UNGER |
Abstract
|
D00164 PC386 00 |
Effective Dose Equivalent for Specific Radionuclides. |
UNIFY-ECN |
Abstract
|
P00288 C0170 00 |
A Program to Calculate Fast Neutron Data for Structural Materials. |
UNIMUG3
|
Abstract
|
C00407 C0170 00 |
Solves Multigroup Diffusion Equations in One-Dimensional Systems. |
UPDATE |
Abstract
|
P00270 DGMV1 00 |
Program to Update Fortran Source Files. |
UPDATE |
Abstract
|
P00270 I3081 00 |
Program to Update Fortran Source Files. |
UPEAK |
Abstract
|
P00300 IPCXT 00 |
A Program for Decomposing A One-Dimensional Spectrum. |
UPEML 3.0 |
Abstract
|
P00245 ALLCP 01 |
A Machine-Portable CDC UPDATE Emulator. |
URR |
Abstract
|
P00281 D6220 00 |
Calculates Resonance Neutron Cross-Section Probability Tables, Bondarenko Self-Shielding Factors and Self-Indication Ratios for Fissile and Fertile Nuclides. |
USINT |
Abstract
|
P00415 MNYCP 00 |
Code System to Calculate Heat and Mass Transfer In Concrete |
USRHYD
|
Abstract
|
C00197 I3675 00 |
Electron and X-Ray Energy Deposition and Hydrodynamics Code System. |
UTMTOX
|
Abstract
|
C00500 D8600 00 |
Unified Transport Model for Toxic Materials. |
UTSG |
Abstract
|
P00379 I3033 00 |
Code System for Calculating the Nonlinear Transient Behavior of a Natural Circulation U-Tube Steam Generator with Its Main Steam System. |
UTXS6 |
Abstract
|
D00211 MNYCP 00 |
MCNP Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1365K. |
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. |
VARSKIN 3 V3.1. 0
|
Abstract
|
C00522 PC586 06 |
Code System for Assessing Skin Dose from Skin Contamination., Version 3.1.0. |
VCS
|
Abstract
|
C00262 I0360 00 |
Coupled Discrete Ordinates-Adjoint Monte Carlo Calculation of Radiation Protection Factors in Vehicles. |
VELM |
Abstract
|
D00133 I0360 00 |
Multigroup Cross-Section Libraries Based on ENDF/B-V Data for Sodium-Cooled Reactor Shield Analysis. |
VENTURE-PC
|
Abstract
|
C00654 PC586 02 |
A Reactor Analysis Code System. |
VIDEO-PC |
Abstract
|
P00311 IBMPC 00 |
Super VGA Primitives Graphics System. |
VIEWCXS |
Abstract
|
P00514 PC586 00 |
Interactive Graphic User Interface to View Neutron and Gamma-Ray Interaction Cross Sections. |
VIM 5.1
|
Abstract
|
C00754 MNYWS 01 |
Continuous Energy Neutron and Photon Transport Code System, April 2009 Release. |
VISA2 |
Abstract
|
P00445 MNYCP 00 |
Code System to Calculate Probability of Reactor Vessel Failure. |
VITAMIN-4C |
Abstract
|
D00053 I3691 00 |
171 Neutron Group Cross Sections and Bondarenko Factors in CCCC Interface Formats for Fusion and LMFBR Neutronics. |
VITAMIN-B6 |
Abstract
|
D00184 ALLCP 00 |
A Fine-Group Cross Section Library Based on ENDF/B-VI Release 3 for Radiation Transport Applications. |
VITAMIN-C |
Abstract
|
D00041 I0360 02 |
171 Neutron, 36 Gamma-Ray Group Cross Sections in AMPX and CCCC Interface Formats for Fusion and LMFBR Neutronics. |
VITAMIN-E |
Abstract
|
D00113 I3033 02 |
174n, 38g Cross-Section Library in AMPX Format. |
VITAMIN-J/COVA |
Abstract
|
D00157 D8810 00 |
Neutron Cross-Section Covariance Data in Multigroup Form. |
VITAMIN-J/COVA/EFF |
Abstract
|
D00197 ALLCP 00 |
Neutron Cross-Section Covariance Data in Multigroup Form. |
VITAMIN-J/KERMA |
Abstract
|
D00150 I3090 00 |
VITAMIN-J 175-Neutron and 38-Photon Kerma And Gas Production Cross Sections. |
VITENEA-J |
Abstract
|
D00238 MNYCP 00 |
AMPX 175-n,42-g Multigroup X-section Library for Nuclear Fusion Applications. |
VIXEN |
Abstract
|
P00030 C6600 00 |
A Code to Check Physical Consistency of Photon-Production Data in Revised ENDF Format. |
VIXEN |
Abstract
|
P00030 I0360 00 |
A Code to Check Physical Consistency of Photon-Production Data in Revised ENDF Format. |
VPI-NECM
|
Abstract
|
C00481 C0740 00 |
Nuclear Engineering Computer Models for In-Core Fuel Management Analysis. |
VPI-NECM
|
Abstract
|
C00481 D0VAX 00 |
Nuclear Engineering Computer Models for In-Core Fuel Management Analysis. |
VPI-NECM
|
Abstract
|
C00481 PC486 00 |
Nuclear Engineering Computer Models for In-Core Fuel Management Analysis. |
VSOP94
|
Abstract
|
C00670 MNYWS 00 |
Code System for Reactor Physics and Fuel Cycle Simulation. |
VVER-BENCHMARKS
| Abstract
|
M00003 MNYCP 00 |
Collection of Neutronic VVER Reactor Benchmarks. |
W-M-NRSM |
Abstract
|
D00026 U1108 00 |
WANL-MSFC Nuclear Rocket Shielding Methods Data Generator (GAMLEG-W, APPROPOS, NAGS, and SATURN) and Multigroup Neutron and Gamma-ray Cross Section Libraries 1-6. |
WEERIE
|
Abstract
|
C00426 I3033 00 |
Code System for Assessing the Radiological Consequences of Airborne Effluents from Nuclear Installations. |
WHATIF-AQ
|
Abstract
|
C00561 B7800 00 |
A Computer Program For Speciation Calculation. |
WILIT |
Abstract
|
P00344 MNYCP 00 |
A Utility Program for WIMS Libraries. |
WIMKAL-88 |
Abstract
|
D00193 MNYCP 00 |
69 Energy Group, Neutron Cross Section Library For Thermal Reactor Calculations in WIMSD Format. |
WIMS-ANL 4.0
|
Abstract
|
C00698 MNYCP 00 |
Deterministic Code System for Reactor Lattice Calculation. |
WIMSCORE-ENEA |
Abstract
|
P00319 I3090 00 |
Code System to Process WIMSD4 Interface Output Files and Generate Two-Group Data for Reactor Calculations. |
WIMSD-5B.12
|
Abstract
|
C00656 MNYCP 02 |
Deterministic Code System for Reactor Lattice Calculation |
WIMSLIB-IJS0 |
Abstract
|
D00147 D8810 00 |
Extended Version of the WIMS 69-group Library. |
WIMSLIB-IJS1 |
Abstract
|
D00147 D8810 01 |
Extended Version of the WIMS 69-group Library. |
WIMSLIB-JEF87 |
Abstract
|
D00095 D0VAX 00 |
JEF-1 Based 69 Group Neutron Data Library. |
WINDOWS |
Abstract
|
P00136 I0360 00 |
A Program for the Analysis of Spectral Data Foil Activation Measurements. |
WINDOWS II |
Abstract
|
P00161 I0370 00 |
A Program for the Analysis of Spectral Data Foil-Activation Measurements. |
WLUP 3.0 |
Abstract
|
D00231 MNYCP 00 |
69- and 172- Group Cross Section Libraries for WIMS. |
WRAITH
|
Abstract
|
C00427 U1100 00 |
Code System for Calculating Internal and External Doses Resulting from an Atmospheric Release of Radioactive Material. |
WREM-TOODEE2 |
Abstract
|
P00469 ALLMF 00 |
2-D Time-Dependent Fuel Element, Thermal Analysis Code System. |
X4ECS |
Abstract
|
P00220 D0780 00 |
A Code System to Combine Cross Section Data in EXFOR and/or ENDF/B-IV Format. |
X4R |
Abstract
|
P00222 DVX11 00 |
Code System for Retrieving EXFOR Cross Section Data According to a Given Target Nucleus. |
XCOM |
Abstract
|
D00174 IBMPC 00 |
Photon Cross Sections on a Personal Computer, Versions 1.2 and 1.3. |
XG-IAEA |
Abstract
|
D00163 IBMPC 00 |
X-ray and Gamma-ray Standards For Detector Calibration. |
XLACS-IIA |
Abstract
|
P00182 I3033 00 |
A Modified Version of XLACS-II for Processing ENDF Data into Multigroup Neutron Cross Sections in AMPX Master Library Format. |
XOQDOQ-82
|
Abstract
|
C00316 DGMV1 00 |
Radiological Assessment Code System - Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations. |
XOQDOQ-82
|
Abstract
|
C00316 I3033 00 |
Radiological Assessment Code System - Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations. |
XOQDOQ-82
|
Abstract
|
C00316 IPCAT 00 |
Radiological Assessment Code System - Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations. |
XPORT-PC
|
Abstract
|
C00559 IBMPC 00 |
An Approximation For Black Body X-Ray Transport in Air. |
XRAY_AAC
|
Abstract
|
C00525 D0750 00 |
X-ray Attenuation and Absorption Calculations. |
XSDRN
|
Abstract
|
C00123 C0073 00 |
Multigroup One-Dimensional Discrete Ordinates Spectral Averaging N Transport Code System. |
XSDRN
|
Abstract
|
C00123 I0360 00 |
Multigroup One-Dimensional Discrete Ordinates Spectral Averaging N Transport Code System. |
XSHLD
|
Abstract
|
C00495 IBMPC 00 |
Diagnostic X-Ray Shielding Calculation. |
YUMMY |
Abstract
|
D00221 MNYCP 00 |
Multi-temperature, Neutron Cross Section Library Based on ENDF/B-V and ENDF/B-VI for use with MCNP. |
ZOTT99 |
Abstract
|
P00272 ALLCP 02 |
Zero-in On The Truth; Evaluation of Correlated Data Using Partitioned Least Squares. |
ZYLIND-PC
|
Abstract
|
C00557 IBMPC 00 |
An Interactive Point Kernel Program For Photon Dose Rate Prediction of Cylindrical Source/Shield Arrangements. |
