Computer Programs

nesc0374 | 1-DX, 1-D Diffusion for Fast Reactor MultiGroup Cross-Sections, Group Constant Collapsing |

nesc0325 | 2-DB, 2-D MultiGroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search |

nea-1250 | 2D-SEEP, 2-D Ground Water Flow in Permeable Geologic Media |

nesc0806 | 2DEPEP, Partial Differencial Equation Solution and Eigenvalues for Potential and Diffusion Problems |

nesc9739 | 2DFLOW, 2-D Drainage Winds and Diffusion Simulation |

iaea1386 | 2GWIHLIB, Generation and Plot of Cross Sections for HYDMN |

nesc0567 | 3-DB, 3-D MultiGroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup |

nea-1250 | 3D-SEEP, 3-D Ground Water Flow in Permeable Geologic Media |

nea-1732 | 3D-TRANS-2003, Workshop on Common Tools and Interfaces for Radiation Transport Codes |

psr-0248 | ABAREX, Optical Statistical Model Neutron Cross-Sections Using ABACUS and NEARREX |

nea-0912 | ABLEIT-TRANS, Isotope Concentration and Sensitivities on Cross-Sections Data |

nea-1839 | ACAB-2008, ACtivation ABacus Code |

nea-0976 | ACCULIB, Program Library of Mathematical Routines |

ccc-0442 | ACDOS3, Neutron Activation Activities and Dose Rates |

nea-1072 | ACFA, Isotope Activation of Coolant and Structure Materials |

iaea0975 | ACORNS, Covariance and Correlation Matrix Diagonalization |

nea-0621 | ACRO, Organ Doses from Acute or Chronic Radioactive Inhalation or Ingestion |

ccc-0372 | ACT-ARA, Time-Dependent Radiation Source Terms |

nea-0511 | ACTIV, Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment |

iaea0960 | ACTIV-JINR, Experimental Gamma Spectra Unfolding |

iaea1380 | ACTIVATE2010, Activation Cross Section by Combining Cross Section and Multiplier (ENDF Format) |

ests0171 | ADASAGE, ADA Application Development System |

nea-0480 | ADDELT, Scattering Law Correlation for Delta Function Phonon Spectra |

nea-1708 | ADEFTA 4.1, Atomic Densities for Transport Analysis |

psr-0190 | ADENA, Fission Products Beta Spectra and Gamma Spectra in 19 Group from U235 Pu239 Mixture |

ccc-0831 | ADVANTG 3.0.3, AutomateD VAriaNce reducTion Generator |

nesc0908 | AERIN, Organ and Tissue Doses from Radioactive Aerosols |

ests0165 | AES, Automated Construction Cost Estimation System |

ccc-0360 | AIRDIF, Neutron and Gamma Doses from Nuclear Explosion by 2-D Air Diffusion |

nea-0001 | AIREK-MOD, Time Dependent Reactor Kinetics with Feedback Differential Equation |

nea-0002 | AIREK-PUL, Periodic Kinetics Problems of Pulsed Reactors |

iaea1274 | AIREKMOD-RR, Reactivity Transients in Nuclear Research Reactors |

nea-1130 | AIRGAMMA, External Gamma-Ray Exposure from Radioactive Cloud |

nesc0326 | AIROS-2A, Space-Independent Reactor Kinetics and Space-Dependent Heat Transfer, Mass Transfer |

ccc-0341 | AIRSCAT, Dose Rate from Gamma Air Scattering by Single Scattering Approximation |

ccc-0110 | AIRTRANS, Time-Dependent, Energy Dependent 3-D Neutron Transport, Gamma Transport in Air by Monte-Carlo |

nea-0590 | AKIMA'S-SPLINE, Curve and Surface Fit of Uni-Variate and Bi-Variate Function |

iaea1432 | AL-SHIELDER, calculates shielding thickness of aluminum for any photon emitting radionuclide between 0.5 to 10 MeV |

nea-0500 | ALARM, Thermohydraulics of BWR with Jet Pumps During LOCA |

nea-0705 | ALARM-P1, PWR Thermohydraulics for ECCS During Blowdown |

nea-1353 | ALBEDO ALBEZ, Gamma and Neutron Attenuation in Air Ducts |

nea-0108 | ALCI, Homogeneous 2-D MultiGroup Neutron Diffusion in X-Z, R-Z, R-Theta Geometry with Criticality Search |

ccc-0577 | ALDOSE, Dose Rate from Alpha Disk Source in H20 |

uscd1238 | ALICE2011, Particle Spectra from HMS precompound Nucleus Decay |

psr-0550 | ALICE2017, Statistical Model Code System to Calculate Particle Spectra from HMS Precompound Nucleus Decay |

ccc-0558 | ALKASYS, Rankine-Cycle Space Nuclear Power System |

ccc-0612 | ALPHN, (Alpha, N) Neutron Production in High-Level Waste Canisters |

nea-0585 | ALPS, Solid-State Detector Alpha Spectra Unfolding |

nesc0815 | ALVIN, Diffusion and Integral Data Comparison and Sensitivity Analysis |

nea-0675 | AMALTHEE, Emission Spectra for N, D, H3, He3, He4 Induced Reactions |

nea-0403 | AMARA, Correlation of Nuclear Data to Integral Experiment by Lagrange Multipliers |

nesc0562 | AMDLIBAE, IBM 360 Subroutine Library, Special Function, Polynomials, Differential Equation |

nesc0563 | AMDLIBF, IBM 360 Subroutine Library, Eigenvalues, Eigenvectors, Matrix Inversion |

nesc0564 | AMDLIBGZ, IBM 360 Subroutine Library for Data Processing, Graphics, Sorting |

iaea1251 | AMICO, Cross-Sections Data for ANISN, DOT from WIMS Library |

ccc-0793 | AMP, Advanced Multi-Physics |

psr-0315 | AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5 |

uscd0795 | AMRAW, Risk Assessment Method for Radioactive Waste Management |

nea-1235 | AND, Atomic Number Densities for Criticality Calculation |

nea-0321 | ANDROMEDA, 1-D Burnup for Fuel Cycle Analysis of FBR |

nea-1798 | ANGELO-LAMBDA, Covariance matrix interpolation and mathematical verification |

nea-0633 | ANIPLO-D50, Plot of Scalar Flux and Dose Rates from ANISN Calculation |

ccc-0082 | ANISN-E, 1-D Transport Program ANISN with Exponential Model |

nea-0363 | ANISN-FONTENAY, 1-D Planar, Spherical, Cylindrical Neutron Transport and Gamma Transport with Deep Penetration |

ccc-0082 | ANISN-JR, 1-D Transport Program ANISN with ZZ JSD Data and Flux Plot |

ccc-0254 | ANISN-ORNL, 1-D Neutron Transport & Gamma Transport in Slab, Cylindrical, Spherical Geometry with Anisotropic Scattering |

ccc-0255 | ANISN-W, 1-D Transport Calculation for Deep Penetration Problems |

ccc-0514 | ANISN/PC, MultiGroup 1-D Discrete Ordinates Transport with Anisotropic Scattering |

nea-1638 | ANITA-IEAF, Isotope Inventories from Intermediate Energy Neutron Irradiation for Fusion Applications |

nea-0470 | ANSCLAD-1, Creep Strain in Fuel Pin Zircaloy Clad During Temperature Transient |

nesc0529 | ANVENT, Temperature Distribution and Pressure in Containment and Ice Condenser after LOCA for LWR |

nesc9977 | ANYOLS, Least Square Fit by Stepwise Regression |

nea-0546 | APPLE, Plot of 1-D MultiGroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN |

nea-0367 | APPROX, 1-D and 2-D Function Approximation by Polynomials, Splines, Finite Elements Method |

nea-0445 | APS-2, Elastic Behaviour of Piping System |

psr-0065 | APSAI, Activation Calculation and Plot of Neutron Spectra, Gamma Spectra by ANISN |

iaea1219 | APUD-3.0, Off-Site Contamination Assessment from Accidental Release |

ests1169 | ARCON96, Radioactive Plume Concentration in Reactor Control Rooms |

nea-0320 | ARGO, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry from JAERI Fast-Set, ABBN, RCBN |

nesc0152 | ARGUS, Transient Temperature Distribution Cylindrical Geometry, Space-Dependent or Time-Dependent Heat Generator |

nea-1368 | ARIANNA-2, Sub-Compartment Thermo-Hydraulic Transients in LOCA |

nea-0174 | ARLEKIN, General Point Reactor Kinetics by Lie-Series Method |

nesc0925 | ARRRG/FOOD, Doses from Radioactive Release to Food Chain |

nesc0738 | ARSTEC, Nonlinear Optimization Program Using Random Search Method |

nea-1581 | ART MOD2, Fission Product Migration in Primary System and Containment |

nea-0539 | ASCOT-1, Thermohydraulics of Axisymmetric PWR Core with Homogeneous Flow During LOCA |

nea-0661 | ASNBILD, Generator of JCL and Data for Program ANISN on CDC Computer |

ccc-0126 | ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport |

ccc-0417 | AT123D, 1-D, 2-D, 3-D Transient Waste Transport Simulation in Groundwater |

psr-0431 | ATHENA_2D, Simulation Hypothetical Recriticality Accident in a Thermal Neutron Spectrum |

ccc-0179 | ATR, Radiation Transport Models in Atmosphere at Various Altitudes |

iaea0906 | AUJP, Optical Potential Parameter Search by CHI**2 Method |

ccc-0519 | AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors |

psr-0008 | AUTOJOM, Quadratic Equation Coefficient for Conic Volume, Parallelepipeds, Wedges, Pyramids |

nea-1076 | AVACOM-ETAP, Availability and Element Transient and Asymptotic Repair Process |

nesc9700 | AVPROG, Monte-Carlo Simulation of System Availability |

nea-0861 | AWE-1 AWE-2 BRUNA, Minimal Cut Sets of Logic Trees |

nesc0191 | AX-TNT, Super Prompt Critical Excursions in Spherical Geometry, Thermohydraulics |

nea-0179 | AXIFLUX, Cosine Function Fit of Experimental Axial Flux in Cylindrical Reactor |

psr-0075 | AXMIX, ANISN Cross-Sections Mixing, Transport Corrections, Data Library Management |

psr-0297 | AXMIX-PC, Cross-Sections Generator for ANISN, DOT from Different Sources |

nesc9564 | AYER, 2-D Thermal Conduction by Finite Element Method |

nesc1020 | BACFIRE, Minimal Cut Sets Common Cause Failure Fault Tree Analysis |

uscd1158 | BALANCE, Mass Transfer in Groundwater Aqueous Solution |

nesc9677 | BARMOM, Fission Barriers and Moments of Inertia |

iaea0953 | BASACF, Integral Neutron Spectra Adjustment and Dosimetry |

nea-0636 | BASKER, Isotropic Scattering Kernel Calculation Using VIWI |

uscd1040 | BAYESZ, S-Wave, P-Wave Resonance Level Spacing and Strength Functions |

iaea1326 | BCS-COLL, Nuclear Level Densities of Excited Nuclei |

iaea0827 | BEAT, Reactor Response and Reactivity Analysis |

nea-0949 | BERMUDA, 1-D, 2-D, 3-D Neutron and Gamma Transport for Shielding |

nea-0373 | BEST-4, Fuel Cycle and Cost Optimization for Discrete Power Levels |

nea-0404 | BEST-5, Power Reactor Fuel Cycle Optimization by Bellman Method |

ccc-0117 | BETA-2B, Time-Dependent Bremsstrahlung Transport, Electron Transport by Monte-Carlo Method |

ccc-0657 | BETA-S, Multi-Group Beta-Ray Spectra |

nea-0591 | BEVE, Isotope Buildup in LWR Fuel Pin with Self-Shielding in Pellet |

nea-0541 | BICUSP, Solution and Derivatives of 2-D Function in Rectangular Mesh Grid by Splines |

nea-0188 | BIGGI-4T, Gamma Transport in Multi-Region Shield in Planar or Spherical Geometry |

ests0298 | BIMOND3, Monotone Bivariate Interpolation |

nesc1037 | BIMOND3, Monotone Bivariate Interpolation |

psr-0117 | BINX, MINX Utility and SPHINX Utility, BCD to BIN Library Conversion |

nea-0870 | BISON, 1-D Burnup and Transport in Slab, Cylindrical, Spherical Geometry |

iaea0820 | BLAST, Accident Conditions in Critical and Subcritical Thermal Reactor System |

psr-0377 | BLOCKAGE2.5R, Plug of Emergency Core Cooling Suction Strainers by Debris BWR |

nea-0683 | BLOK, Turbulent Flow in Pipes and Channels with Rectangular Obstruction |

nea-0978 | BLOOM, Principal Component Analysis and Correspondence Analysis Using IMSL Subroutines |

ccc-0633 | BLT, Waste Transport through Porous Media from Container Failure |

nea-0660 | BOB-7, Ge(Li) Detector Gamma Spectra Unfolding |

ccc-0459 | BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup |

nea-0236 | BOLERO, 2 Group Burnup for PWR and BWR in R-Z Geometry with Restart and Recycle |

iaea1246 | BOMJ, Level Assignments from Gamma Spectra Measurements |

psr-0173 | BON, Unfolding of Multisphere Spectrometer Neutron Spectra |

nea-1678 | BOT3P5.3, 3D Mesh Generator and Graphical Display of Geometry for Radiation Transport Codes, Display of Results |

nea-1523 | BOXER, Fine-flux cross section condensation, 2D few group diffusion and transport burnup calculations |

iaea0915 | BRA, Breast Radiation Analysis from Mammography |

nea-0516 | BRANCALEONE, Transfer Function Roots for Linear System of Several Variables |

psr-0143 | BREESE, Distribution Function for Program MORSE from Albedo Data |

iaea1190 | BRETISLAV,OLDRICH, Neutron Diffusion in Hexagonal Geometry for WWER Critical Fuel Assemblies |

nesc9804 | BRGLM, Interactive Linear Regression Analysis by Least Square Fit |

nea-0390 | BRIGITTE, Dose Rate and Heat Source and Energy Flux for Self-Absorbing Rods |

nea-0438 | BRIGITTE-KA, ENDF/B to KEDAK Data Conversion with Resonance Cross-Sections Tables Generator |

nea-0418 | BRUCH-D-06, LOCA of PWR Primary System with 23 Control Volume and 9 Rupture Points |

nea-0866 | BTPLOT BTSPEC EXSPEC ORDTAB TABLST, Retrieval of ENDF/B Decay Spectra |

nesc0667 | BUCKLE, Time-Dependent Deformation of 1-D Oval Pipe Under Pressure, Temperature, Neutron Flux |

nea-1727 | BULK-I, Radiation Shielding Tool for Proton Accelerator Facilities |

nea-1771 | BULK_C-12, N & photon effective dose rates from medium energy protons or carbon ions through concrete or concrete/iron |

nea-1819 | BURD, Bayesian estimation in data analysis of Probabilistic Safety Assessment |

nea-0237 | BURNY, 5 Group BWR and PWR Burnup in X-Y Geometry by Diffusion Calculation |

nea-0350 | BURNY-SQUID, 2-D Burnup of UO2 and Mix UO2 PuO2 Fuel in X-Y or R-Z Geometry |

nea-0114 | BURST, Time-Dependent Pressure and Coolant Flow after Circuit Fracture in HTGR |

nesc0435 | BURST-1, Rupture of 1-D Cylindrical Pressurized Liquid System, Hydrodynamic Calculation |

nea-0558 | BUST, Elastic Stress in HTGR Pressurized Fuel Elements |

nea-0159 | BWCAL, Void Distribution and Flow Velocity in BWR |

uscd1151 | BWIP-RANDOM-SAMPLING, Random Sample Generation for Nuclear Waste Disposal |

nesc1080 | BWR-GALE, Radioactive Gaseous and Liquid Waste Release from BWR |

ccc-0485 | BWR-LTAS, BWR Long Term Accident Simulation Program |

nea-1313 | BWRDYN, Thermal Hydraulic Analysis of a BWR Plant |

nea-1044 | BWRPLANT/ZERO, Dynamic Model for BWR Nuclear Plant |

iaea1403 | C-SHIELDER, Gamma shielding calculations of radionuclides emitting photons 0.5 to 10 MeV by different concretes |

ccc-0476 | CAAC, System to Implement Atmospheric Dispersion Assessments |

nea-1020 | CADE, Multiple Particle Emission Cross-Sections by Weisskopf-Ewing Theory |

nesc0270 | CAESAR-4, 1-D MultiGroup Diffusion in Slab, Cylindrical, Slab Geometry, Criticality Search |

nea-1800 | CAFDATS, Converter of Angular Fluxes of DORT, ANISN and TORT Systems |

nea-1278 | CALENDF-2010, Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations |

ccc-0594 | CALKUX, Exposure Transmission of Medical X-Ray Beams Through Barrier Materials |

ccc-0610 | CALOR95, High-Energy Calorimeter Design and Data Evaluation by Monte-Carlo |

ccc-0240 | CAMERA CAM, Radiation Dose Absorption by Computer Man |

ccc-0542 | CAP-88, Dose Risk Assessment from Air Emissions of Radionuclides |

nea-1327 | CAPCAL, 3-D Capacitance Calculator for VLSI Purposes |

nea-0290 | CARBOX, Equilibrium of Non-Stoichiometric Mixtures of Oxides, Carbides, Methane |

nesc0638 | CAREN-4, ENDF/B Utility, Discontinuity Check at Resonance Region Boundary |

psr-0388 | CARES, Seismic Structure Safety Analysis for Nuclear Power Plants |

ests0012 | CARES-ESTSC, Seismic Structure Safety Analysis for Nuclear Power Plants |

nea-1735 | CARL 2.3, radiotoxicity, activity, dose and decay power calculations for spent fuel |

nea-0649 | CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback |

nea-0393 | CARNAC, Neutron Flux and Neutron Spectra in Criticality Accident |

psr-0131 | CARP-82, Multigroup Albedo Data using DOT Angular Flux Results |

ccc-0024 | CARSTEP, Particle Flux on Space Vehicle in Van Allen Zone |

nesc0482 | CASCADE, Intranuclear Gamma Cascade Calculation for Particle Emission Probability |

psr-0262 | CASKCODES, Program CAPSIZE Scope KWIKDOSE for Shipping Cask Shielding |

nea-1195 | CASKET, Thermal and Structural Analyses for Transport and Storage Cask |

nea-0712 | CASSANDRE, 2-D Reactor Dynamic FEM Program with Thermohydraulic Feedback |

nea-1395 | CASTHY, Statistical Model for Neutron Cross-Sections and Gamma-Ray Spectra |

iaea1347 | CCRMN, N, P, He4, D, H3, He3 Reaction Calculation for Medium-Heavy Targets |

nesc9789 | CDMS, Cost Data Management System Spread Sheet |

iaea0920 | CEBIS, 1-D 2 Group Diffusion Code for Reactor Calculation |

ests1071 | CECP, Decommissioning Costs for PWR and BWR |

nea-0553 | CEDRAZAL, Steady-State Heat Transfer in HTR with Multifuel Region |

psr-0532 | CEM03.03, Monte-Carlo Code system to calculate nuclear reactions in the framework of the improved cascade-exciton model |

iaea1247 | CEM95, Cascade Exciton Model Nuclear Reactions by Monte-Carlo Method |

ccc-0544 | CEPXS ONELD, 1-D Coupled Electron Photon MultiGroup System |

ccc-0837 | CEPXS, Coupled Electron-Photon Cross Section |

nea-0648 | CERBERO, Cross-Sections by Optical, Statistical Model for Spin 0, Spin 1/2 Particles |

nesc0415 | CEXE INCEXE, 1 Group 3-D Time-Dependent Xe Oscillations in X-Y-Z Geometry with Feedback |

ests0663 | CFDLIB, Computational Fluid Dynamics Library |

nesc9537 | CFEST-1.1, Coupled Fluid, Energy, Solute Transport in Ground-Water System |

iaea1266 | CFUP1, Neutron or Charged-Particle Reactions of Fissile Nuclei up to 33 MeV |

ccc-0604 | CHAINS-PC, Decay Chain Atomic Densities |

ccc-0584 | CHAINT-MC, 2-D Radionuclide Transport in Fractured Porous Medium |

ccc-0070 | CHARGE-2/C, Flux and Dose Behind Shield from Electron, Proton, Heavy Particle Irradiation |

nesc0638 | CHECK-4, ENDF/B Utility, Structure Consistency Check and Format Check |

uscd1208 | CHECKR, ENDF/B Format Check |

nea-1561 | CHEMENGL/CHIMISTE, Chemical and Physical Properties of Elements |

nea-1346 | CHEMTARD, Simulation of Chemical Species Through Porous Media |

nea-0716 | CHOLESK, Diffusion Calculation with 2-D Source in X-Y or R-Z Geometry |

uscd1021 | CHUCK-3, Nuclear Scattering Amplitude and Collision Cross-Sections by Coupled Channel |

nea-0451 | CICLON, Neutronics Calculation for PWR Transition Fuel Cycle Management |

ccc-0755 | CINDER 1.05, Actinide Transmutation Calculations Code |

nesc0313 | CINDER, Depletion and Decay Chain Calculation for Fission Products in Thermal Reactors |

psr-0117 | CINX, MINX Utility and SPHINX Utility, Library Data Collapsing |

nesc9602 | CIRCLE-SPLINE, 2-D, 3-D Spline Curve Fitting |

nesc0387 | CITATION, 3-D MultiGroup Diffusion with 1st Order Perturbation and Criticality Search |

ccc-0643 | CITATION-LDI2, 2-D MultiGroup Diffusion, Perturbation, Criticality Search, for PC |

iaea1385 | CITOPP,CITMOD,CITWI, Processing codes for CITATION Code |

nea-0631 | CLAPTRAP, Pressure Transients in LWR Containment During LOCA |

iaea0883 | CLUB, Cell Calculation PF Candu PWR Fuel Clusters |

nea-0864 | CLUHET, Steady-State Thermohydraulics of Rod Bundles with 1 Phase Flow |

nea-0357 | CLUP-77C, Collision Probability and Neutron Flux Distribution in Multi-Region Fuel Cluster |

nea-0255 | CLUS, Heat Transfer and Fuel Power in Liquid Cooled 7 Rod Fuel Elements Cluster |

iaea1265 | CMUP2, Reaction Cross-Sections for N, P, D, T, He3, He4 up to 50 MeV |

ccc-0726 | CNCSN 2009, One, Two- and Three-Dimensional Coupled Neutral and Charged Particle Sn Parallel Multi-Threaded Code System |

nesc0873 | COAST-4, Design and Cost of Tokamak Fusion Reactors |

nesc9978 | COBRA-3C/RERTR, Thermohydraulic Low Pressure Subchannel Transients Analysis |

nesc0432 | COBRA-4I, Transient Thermohydraulics Fuel Elements Clusters, Subchannel Analysis Method |

nea-1614 | COBRA-EN, Thermal-Hydraulic Transient Analysis of Reactor Cores |

psr-0614 | COBRA-SFS CYCLE 4A, Code System for Thermal Hydraulic Analysis of Spent Fuel Casks |

ests0135 | COBRA-SFS CYCLE3, Thermal Hydraulic Analysis of Spent Fuel Casks |

nesc1091 | COBRA-SFS, Thermal Hydraulics of Spent Fuel Storage System |

nea-0294 | CODAC, MultiGroup Cross-Sections Generation from ENDF/B for Monte-Carlo Program TIMOC |

ccc-0829 | COG11.1, Multiparticle Monte Carlo Code System for Shielding and Criticality Use |

psr-0375 | COGAP, Nuclear Power Plant Containment Hydrogen Control System Evaluation Code |

nea-0915 | COGEND, Decay Data Generated in ENDF-6 Format |

psr-0607 | COGLIBMAKER2014, Data Conversion Utility |

nea-1126 | COLLI-PTB, Neutron Fluence Spectra for 3-D Collimator System by Monte-Carlo |

nea-0903 | COLUMN, 1-D Migration for Various Physical Chemical Processes |

psr-0286 | COMBINE-PC, MultiGroup Neutron Cross-Sections in B1 or B3 Approximation from ENDF/B-5 |

nea-0340 | COMET, Mechanical and Thermal Stress in Fuel Element Clad |

psr-0343 | COMIDA, Radionuclide Food Chain Model for Acute Fallout Deposition |

nesc0482 | COMNUC, Gamma Emission Neutron Emission Fission and Scattering Cross-Sections Using Hauser-Feshbach |

psr-0302 | COMNUC3B, Gamma Emission, Neutron Emission Fission and Scattering Cross-Sections Using Hauser-Feshbach |

iaea0966 | COMPAR, NJOY, GROUPIE, FLANGE-2, ETOG-3, XLACS MultiGroup Cross-Sections General Comparison |

nesc0702 | COMPARE, Transient Subcompartment Thermodynamics Analysis with 2 Phase Vent Flow |

nesc0776 | COMPARE-MOD1 COMPARE-MOD1A, 2 Phase Flow Thermodynamics, Pressure in LWR Containment |

ests0023 | COMPBRN3, Modelling of Nuclear Power Plant Compartment Fires |

iaea1321 | COMPLOT2010, Compare ENDF/B Plots of Reaction Data |

nesc0649 | COMQC, Quality Control Statistical Analysis for Means, Errors, Skewness, Kurtosis |

nea-1578 | COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System |

nesc0663 | COMRADEX-4, Doses from Radioactive Release, Meteorological Dispersion, Aerosol |

iaea0928 | COMTA, Ceramic Fuel Elements Stress Analysis |

nesc0498 | CONCEPT-5, Cost and Economics Analysis for Nuclear Fuel or Fossil Fuel Power Plant |

ests0680 | CONCHAS-SPRAY, Reactive Flows with Fuel Sprays |

nea-0325 | CONDENSE, Conversion of JAERI Fast-Set to ABBN Format with Self-Shielding |

nea-0946 | CONDN-63B, Thermohydraulics of Nuclear Power Plant Condenser |

nea-0427 | CONDOR-3, Local and Spectrum Dependent Burnup with Mesh-Wise Depletion |

ccc-0416 | CONDOS-II, Radiation Dose from Consumer Product Distribution Chain |

nesc0433 | CONTEMPT, LWR Containment Pressure and Temperature Distribution in LOCA |

nesc0818 | CONTEMPT-4/MOD5&6, LWR Containment Long-Term Pressure Distribution and Temperature Distribution in LOCA |

iaea1307 | CONVERT2010, FORTRAN Program Converter for Different Computers |

psr-0017 | COOLC, Ne-213 Liquid Scintillation Detector Neutron Spectra Unfolding |

nea-1305 | COOLOD, Steady-State Thermal Hydraulics of Research Reactors |

nea-0567 | CORAN, PWR and BWR Containment Response to LOCA |

iaea1226 | CORD, PWR Core Design and Fuel Management |

nesc0758 | COREL, Ion Implantation in Solids, Range, Straggling Using Thomas-Fermi Cross-Sections |

nesc0759 | CORTES, Steady-State and Transient Heat Flow and Stress Analysis in Pipe Joints |

nea-0383 | COSANI-2, Gamma Doses from SABINE Calculation, Activity from ANISN Flux Calculation |

nea-1375 | COSIMA, BWR Core Performance Simulator |

nea-0160 | COSTANZA-AX, 1-D Neutronics and Thermodynamics of Liquid Cooled Reactor in Axial Geometry |

nea-0160 | COSTANZA-CYL, 1-D Neutronics and Thermodynamics of Liquid Cooled Reactor in Cylindrical Geometry |

nea-0333 | COSTANZA-RZ, 1-D Liquid Cooled Reactor Dynamic in R-Z Geometry |

nea-0425 | COSTANZA-XE, 2-D Pebble-Bed or Prismatic Fuel Elements HTR Dynamic in Cylindrical Geometry |

nea-0398 | COSTAX-BOIL, Transient Dynamic Analysis of BWR and PWR in Axial Geometry |

nea-0533 | COSTAX-BWR, Coupled Time-Dependent 2 Group Neutron Diffusion and 2 Phase Fuel Rod Coolant Flow |

nea-0574 | COVAL, Compound Probability Distribution for Function of Probability Distribution |

ccc-0419 | CRAC2, Reactor Accident Risk Assessment |

nea-0463 | CRACKLE, Fast Reactor Pu Fuel Management |

nea-0057 | CRAM-360, 1-D, 2-D Multi Group Diffusion with Keff Calculation or Criticality Search |

nea-0718 | CRAPONE, Optical Model Potential Fit of Neutron Scattering Data |

nesc0638 | CRECT, ENDF/B Utility, Data Correlation and Data Update |

nea-0948 | CRECT-J, Input Preparation of Evaluated Data in ENDF-4, ENDF-5 and ENDF-6 Formats |

nesc9958 | CREEP-80, Creep Analysis of Concrete Structure by Finite Element Method |

nea-1734 | CRISSUE-S, Neutronics/Thermal-hydraulics Coupling in LWR Technology |

nea-1903 | CRISTAL V2.0.2, Criticality calculation package |

iaea0873 | CRITIC, In-Core Fuel Management for CANDU PWR |

ccc-0518 | CRRIS, Health Risk Assessment from Atmospheric Releases of Radionuclides |

nea-1040 | CRUNCH, Dispersion Model for Continuous Dense Vapour Release in Atmosphere |

ccc-0233 | CRYSTAL-BALL, Neutron Spectra Calculation from Activation Experiment with Error Estimate |

nea-1892 | CUMYIELD.MT, cumulative yields calculations of radioactive decay isotopes considering decay chain |

nea-0507 | CURFIT SURFIT, 2-D Polynomial Least Square Fit to Experimental Data |

nea-0247 | CYGAS, 3-D Gamma Flux in Axial or Cylindrical Shields from Cylindrical Source |

nea-0494 | CYLDOS, Dose Rate in Cylindrical Shield from Cylindrical Source |

nea-0371 | CYLFUX, Fast Reactor Reactivity Transients Simulation in LWR by 2-D 2 Group Diffusion |

nea-1535 | D2O, Computation of Thermodynamic and Transport Properties of Heavy Water |

nea-1416 | D3D/D3E, 2-D, 3-D MultiGroup Neutron Diffusion in Rectangular, Cylindrical, Triangular-Z Geometry |

ccc-0273 | DACRIN, Dose in Respiratory Tract and Organs from Aerosol Inhalation |

nesc0758 | DAMG2, Ion Implantation in Solids, Energy Deposition Distribution with Recoils |

nea-0151 | DANCOFF-3, Dancoff Correction for Cylindrical Fuel Rod at H2O Gaps and for Fuel Clusters |

nea-1516 | DANCOFF-MC, Dancoff Correlation for Arbitrary Lattices by Monte-Carlo |

nea-0103 | DANG, Elastic and Direct Inelastic and Reaction Neutron Cross-Sections, Deformed Even-Even Nuclei |

nea-0694 | DANTE, Activation Analysis Neutron Spectra Unfolding by Covariance Matrix Method |

ccc-0547 | DANTSYS3.0, 1-D, 2-D, 3-D MultiGroup Discrete Ordinate Method Transport |

nea-1885 | DART-V.1, displacement per atom, primary knocked-on atoms produced in an atomic solid target |

ccc-0366 | DASH, Void Tracing Sn and Monte-Carlo Coupling Program with Angular Fluxes from DOT Program |

ests0357 | DASH-FP, Multicomponent Time-Dependent Concentration Diffusion |

nea-0646 | DASQHE, Dancoff Correlation for Infinite Circular Rod Assembly in Square or Hexagonal Lattice |

nesc9918 | DASSL, Solution of Differential Algebraic Equation |

nesc9493 | DATING, Temperature for Spent Fuel Dry Storage |

ccc-0640 | DCHAIN, Isotope Buildup and Isotope Decay by 1 Point Approximation |

nea-0664 | DCHAIN, Isotope Buildup and Isotope Decay by 1 Point Approximation |

nea-1603 | DCHAIN-SP 2001, Code System for Analyzing Decay and Build-up Characteristics of Spallation Products |

ccc-0520 | DCTDOS, Neutron and Gamma Penetration in Composite Duct System |

nea-0229 | DCXE, Time-Dependent Xe Diffusion in Non-Multiplying Slab |

ests0848 | DDASAC, Double-Precision Differential or Algebraic Sensitivity Analysis |

iaea1290 | DDCS, P, D, T, He3, He4 Reactions with 5 Particle Emission by Optical Model |

nesc0640 | DE/STE/INTRP, 1st Order Ordinary Differential Equation for Initial Value Problems |

nea-1893 | DECAYHEAT.MT, decay heat calculations from radioactive isotopes |

nea-0834 | DEEBAR, Resonance Level Spacing Calculation by Dyson-Metha Optimum Statistics |

ccc-0455 | DEIS, Impact Measures of Low Level Radioactive Waste Disposal |

nea-0446 | DELIGHT-7, Point Reactivity Burnup for HTGR Lattice with P1 Neutron Scattering Approximation |

nesc9681 | DEM4-26, Least Square Fit for IBM PC by Deming Method |

nesc0754 | DEMONR, Monte-Carlo Shielding Calculation for Neutron Flux and Neutron Spectra, Teaching Program |

ests0763 | DENDRO, Cluster Analysis of Experimental Data with Tree Plot |

nea-0840 | DENZ, Dense Toxic or Explosive Gases Dispersion in Atmosphere |

nea-0453 | DEPCO-MULTI, Subcooled Decompression in PWR Primary System LOCA |

psr-0523 | DEPLETOR Version 2, provides depletion capability to the Purdue Advanced Reactor Core Simulator (PARCS) code |

nea-1887 | DESAE, develop prospective nuclear energy scenarios in a regional and global scale |

iaea0891 | DIAG, 2-D Plotting Program for PDP-11/34 |

nea-0672 | DIAMANT-2, MultiGroup Neutron Transport with Anisotropic Scattering in Triangular Geometry |

nesc0638 | DICT-4, ENDF/B Utility, Section Table of Contents Generator |

iaea1308 | DICTIN2010, Reaction Index Generated for ENDF Format |

ccc-0649 | DIF3D 8.0/VARIANT8.0, 2-D 3-D Multigroup Diffusion/Transport Theory Nodal & Finite Difference Solver, Variational Method |

ccc-0784 | DIF3D10.0, Variational Nodal Methods, Finite Difference Methods to Solve N diffusion & Transport Theory Problems |

iaea1269 | DIFBAS, Spectra Unfolding of Ne213 P Recoil Detectors |

nea-0667 | DIFFAX, Axial Streaming for Hexagonal Lattices in Gas Cooled FBR, Slab Geometry Diffusion |

nesc0737 | DIFFUSER, 2-D and 3-D Diffuser Performance, Boundary Layer and Turbulent Flow |

nea-0808 | DIFFUSION-ACE, 3-D Neutron Diffusion by Leakage Iteration Method |

nea-1067 | DIFMOD, Radionuclide Leaching and Cement Corrosion in Brine |

nesc9639 | DIGLIB, General Graphics Subroutine Package for Different Computers |

ests0243 | DIGLIB, Multi Platform Graphics Subroutine Library |

nea-0625 | DINE, Neutron Flux, Neutron Dose Rate in Multi-Region Slab Reactor Shield by Removal Diffusion |

nea-0298 | DISCOUNT-G, Nuclear Power Program with Cost Analysis and Pu Production Optimization |

nea-0643 | DISCUS, Neutron Single to Double Scattering Ratio in Inelastic Scattering Experiment by Monte-Carlo |

ccc-0170 | DISDOS, Kerma in Model Man from External Gamma Source |

ccc-0454 | DISPERS, Radioactive Release into Surface Water and Ground Water |

nesc0847 | DISPL-1, 2nd Order Nonlinear Partial Differential Equation System Solution for Kinetics Diffusion Problems |

nesc9532 | DISPOSAL_SITE, Low-Level Radioactive Waste Storage Cost Analysis |

nea-0184 | DIXY-2, 2-D Homogeneous and Inhomogeneous Neutron Diffusion N X-Z, R-Z, R-Theta Geometry with Perturbation |

nea-0391 | DLS, 2-D Diffusion with Line-of-Sight Method for Cavities |

psr-0155 | DOGS, Flux Plots of Radiation Transport Program Using DISSPLA |

psr-0064 | DOMINO, Coupling of Discrete Ordinate Program DOT with Monte-Carlo Program MORSE |

iaea0961 | DOMUS, Experimental 2-D Spectra Analysis |

ccc-0650 | DOORS, 1-,2-,3-dimensional discrete-ordinates system for deep-penetration neutron and photon transport |

psr-0110 | DOQDP ADOQ, Discrete Ordinate Quadrature Generator for Programs DOT and ANISN |

nesc1146 | DORIAN, Bayes Method Plant Age Risk Analysis |

ccc-0543 | DORT, 1-D 2-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration |

ccc-0532 | DORT-PC, 2-D Discrete Ordinates Transport System |

nea-1711 | DORTDAT2, Input-Making Support System for a Two-Dimensional SN Code, DORT |

ccc-0624 | DOSE-SGTR, Iodine Release During Steam Generator Tube Rupture (SGTR) in PWR |

ccc-0536 | DOSEFACTOR-DOE, Dose Rate Conversion Factors for Photon and Electron Exposure |

iaea0922 | DOSKMF2, Dose Rate Distribution in Co60 Gamma Irradiation Plants |

ccc-0276 | DOT-3.5, 2-D Neutron Transport, Gamma Transport Program DOT with New Space-Scaling |

ccc-0320 | DOT-4.2, 2-D Neutron Transport, Gamma Transport with Space Dependent Mesh and Quadrature |

ests0599 | DPCT, Probabilistic Deterministic Contaminant Transport in Ground Water |

nea-1506 | DPOL3D, 2 Group, 3-D Core Transients and Steady State |

uscd1234 | DRAGON 3.05D, Reactor Cell Calculation System with Burnup |

ccc-0647 | DRAGON, Reactor Cell Calculation System with Burnup |

uscd1237 | DRAGON2PARTISN, Cross-Sections Data Generation for PARTISN4.0 |

nea-1412 | DRAWBS, NJOY Graphics Output of ENDF, PENDF, GENDF Data in GKS Format |

iaea0885 | DRUCK, Thermal, Mechanical Stress of PWR Fuel Rod During LOCA Blowdown |

nea-0215 | DRUCKSCHALE-44, Pressure and Temperature Transients in Blowdown Accident |

nea-0839 | DRUFAN-01/MOD2, Transient Thermohydraulics of PWR Primary System LOCA |

nea-0457 | DRUGEVO, Time-Dependent Containment Pressure and Temperature in BWR or PWR LOCA |

ests0637 | DSEM, Radioactive Waste Disposal Site Economic Model |

nesc0784 | DSNP, Program and Data Library System for Dynamic Simulation of Nuclear Power Plant |

psr-0251 | DSNQUAD, Angular Quadrature Weights and Cosines for ANISN |

nesc0209 | DTF-4, 1-D MultiGroup Time-Independent Boltzmann Equation, Slab, Cylindrical, Spherical Geometry, Sn-Method, Pl-Method |

nea-0269 | DTF-G, Reactivity and Flux by 1-D Sn Method on Planar Cylindrical Spherical Geometry |

nea-0322 | DTF4-J, 1-D Neutron Transport with Anisotropic Scattering by Sn Method |

nea-1671 | DUCT-III, Design Code for Duct-Streaming Radiations |

ccc-0453 | DUST, Albedo Monte-Carlo Simulation of Neutron Streaming in Multilegged Square Concrete Ducts |

ccc-0634 | DUST-BNL, Radioactive Waste Transport from Container Leaks into Ground Water |

nesc0579 | DWARF, 1-D Few-Group Neutron Diffusion with Thermal Feedback for Burnup and Xe Oscillation |

nea-1209 | DWBA07/DWBB07, elastic scattering with nucleon-nucleon potential and DWBA for inelastic scattering |

ccc-0383 | DWNWND, Downwind Atmospheric Concentration and Dispersion by Gaussian Plume Model |

nesc9872 | DWUCK-4/5, Scattering Cross-Sections of Spin 0 and 1/2 and 1 Particles by DWBA |

nea-1411 | DYN3D/M2, Reactivity Transients in Light H2O Reactors with Hexagonal Geometry |

nesc0440 | DYNAM, Once Through Boiling Flow with Steam Superheat, Laplace Transformation |

nea-0090 | DYNAMF, Time-Dependent Reactor Dynamics by Laplace Transformation |

nea-0217 | DYNAPS, Vibration Analysis of Piping System in Earthquake |

ests1300 | E3D, 3-D Elastic Seismic Wave Propagation Code |

nea-1813 | EASYQAD 2.0, Visualization for Gamma and Neutron Shielding Calculations |

ests0288 | EBQ, Steady-State Space Charge Transport in Cylindrical Geometry |

nea-0850 | ECIS-12, Coupled Channel, Statistical Model, Schroedinger and Dirac Equation, Dispersion Relation |

ests0219 | ECO2N, a TOUGH2 fluid property module for mixtures of water-NaCl-CO2 |

psr-0191 | EDISTR, Nuclear Data Base Generator for Internal Radiation Dosimetry Calculation |

nea-0969 | EDMULT-6.4, Electron Depth Dose Distribution in Multilayer Slab Absorbers |

nea-0845 | EDO, Doses to Man and Organs from Reactor Operation Noble Gas and Liquid Waste Release |

nea-1028 | EDSPA, 1-D Mechanical Displacement for Elastic, Thermoelastic, Viscoelastic Behaviour |

nesc9575 | EDTGRAF, DISSPLA User Interface Program |

nesc0600 | EGAD, Ground Level Gamma Doses Function of Gamma Energy for Radioactive Releases |

ccc-0331 | EGS4, Electron Photon Shower Simulation by Monte-Carlo |

nesc0983 | EGUN, Charged Particle Trajectories in Electromagnetic Focusing System |

nesc0534 | EISPACK, Subroutines for Eigenvalues, Eigenvectors, Matrix Operations |

ccc-0119 | ELBA, Bremsstrahlung Dose from Isotropic Electron Flux on Plane Al Shield |

nesc0650 | ELBOW, Stress Analysis, Flexibility Factors for Curved Pipes with Internal Pressure |

nea-1200 | ELEORBIT, 3-D Simulation of Electron Orbits in Magnetic Multipole Plasma Source |

ccc-0295 | ELGATL, Calculation of Energy Spectra from Coupled Electron-Photon Slowing Down |

nea-0435 | ELIESE-3, Elastic, Inelastic, Reaction Cross-Sections, Polarization, by Hauser-Feshbach |

iaea1223 | ELPHIC-PC, Statistical Model Monte-Carlo Simulation of Heavy Ion Nuclear Reactions |

ccc-0301 | ELPHO, Muon, Electron, Positron Generator from Pions by Monte-Carlo with HETC Collision Data |

nesc0546 | EMERALD, Radiation Release and Dose after PWR Accident for Design Analysis and Operation Analysis |

nesc0685 | EMERALD-NORMAL, Routine Radiation Release and Dose for PWR Design Analysis and Operation Analysis |

iaea1169 | EMPIRE-II 2.18, Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections |

uscd1235 | ENDF-UTILITY-CODES, codes to check and standardize data in the Evaluated Nuclear Data File (ENDF) |

nea-1895 | ENDSAM, a code for random sampling and validation of covariance data of resonance parameters in ENDF-6 format |

iaea1402 | ENDVER-ENDVER/GUI, The ENDF File Verification Support Package |

uscd1149 | ENSDF ADDGAM, Adds Gammas to Adopted Data Sets |

uscd1149 | ENSDF ALPHAD, Calculation Alpha Hindrance Factors |

uscd1149 | ENSDF AVETOOLS, Three statistical methods to calculate averages of experimental data with uncertainties |

uscd1149 | ENSDF BRICC, Interpolates Band-Raman internal conversion and electron-positron pair coefficients and E0 form factors |

uscd1149 | ENSDF DELTA, Gamma-Gamma Correlation Analysis |

uscd1149 | ENSDF ENSDAT, Graphics and Tables Generation from ENSDF Data |

uscd1149 | ENSDF FETCH, Indexing of ENSDF Files |

uscd1149 | ENSDF FMTCHK, Format Checking Program |

uscd1149 | ENSDF GABS, Absolute Gamma-Ray Intensities from ENSDF Data |

uscd1149 | ENSDF GTOL, Least Squres Fit of Gamma Spectra and Level Assignment |

uscd1149 | ENSDF HSICC, Interpolation Between Hager-Seltzer and Dragoun-Plajner-Schmutzler |

uscd1149 | ENSDF LOGFT, Beta-Decay log-ft and Partial Capture Calculation |

uscd1149 | ENSDF MEDLIST, Dose Rates from Nuclear Decay Data (X-ray intensities) |

uscd1149 | ENSDF NSDFLIB, Subroutine Library for ENSDF Programs |

uscd1149 | ENSDF PANDORA, Physics Checks on ENSDF Data |

uscd1149 | ENSDF PROCESSING CODES, Analysis and Utility Programs |

uscd1149 | ENSDF RADLST, Dose Rates from Nuclear Decay Data (decay of nuclei) |

uscd1149 | ENSDF RULER, Reduced Transition Problems Abilities Calculation |

uscd1149 | ENSDF SPINOZA, Tables of Levels, Decay, Gammy-Ray Data from ENSDF |

uscd1149 | ENSDF TREND, Tabulation of ENSDF Data |

nea-0817 | ENTOSAN, 640 Group Constant Calculation with Resonance from ENDF/B |

nea-1686 | ENTREE 1.4.0, BWR Core Simulation System for Space and Time Dependent Coupled Phenomena |

iaea1285 | EPICSHOW, Interactive Viewing of EPIC (Electron Photon Interaction Code) Data Library |

nesc0675 | EPISODE, 1st Order Stiff or Non-Stiff Ordinary Differential Equation, Initial Value Problems |

nesc0705 | EPISODE-B, 1st Order Stiff or Non-Stiff Ordinary Differential Equation, Initial Value Problems |

nesc0886 | EQ-3 EQ6, Thermodynamics Equilibrium for Aqueous Solution Mineral System |

iaea1202 | EQUIVA, Few-Group Diffusion Parameter for PWR Reflector Region by 1-D Transport Calculation |

nea-0261 | EQUSTA, Thermodynamics Analysis and Mechanical Analysis for Fast Reactor Accident |

nea-1683 | ERANOS 2.3, Modular code and data system for fast reactor neutronics analyses |

nea-0458 | ERDBEBEN, Structure Displacements and Forces Under Earthquake Conditions |

nea-0534 | EREBUS, Burnup by 2-D MultiGroup Neutron Diffusion with Criticality Search |

nea-0815 | ERINNI, Emission Spectra for Multiple Cascades by Optical Model |

nea-0515 | EROS-2, Time-Dependent of Linear System by Inverse Laplace Transformation |

nea-1676 | ERRORJ, Multigroup covariance matrices generation from ENDF-6 format |

nea-0341 | ERUPT, 2-D 2 Group Fuel Management in R-Z Geometry with Fuel Shuffling |

nea-0561 | ESDORA, Continuous and Instantaneous Fission Products Release into Atmosphere |

iaea1282 | ESTAR PSTAR ASTAR, Stopping Power and Range of Electrons, Protons, Alpha |

nea-0892 | ESTIMA, Neutron Width Level Spacing, Neutron Strength Function of S- Wave, P-Wave Resonances |

nea-0984 | ETHEL, Thermos Cross-Sections Library Generator Program |

nea-0394 | ETOA, ABBN MultiGroup Constants from ENDF/B for Fast Reactors |

nea-1048 | ETOBOX, Cross-Sections Library Generated from ENDF/B for Program BOXER |

nesc0350 | ETOE ETOE-2, Cross-Sections Library for Program MC**2 Generator from ENDF/B |

nea-0630 | ETOI, Format Conversion of Resonance Parameter from ENDF/B to Program IRESINT-3 Library |

ccc-0107 | ETRAN, Electron Transport and Gamma Transport with Secondary Radiation in Slab by Monte-Carlo |

nea-0408 | EURCYL, Mesh Generator for 3-D Intersections of Pressure Vessel Nozzles |

nea-1094 | EURDYN, Nonlinear Transient Analysis of Structure with Dynamic Loads |

nea-0447 | EUREKA, Reactivity Transients in LWR from Control Rod, Coolant Flow, Temperature |

iaea1322 | EVALPLOT2010, ENDF Plots Cross Section, Angular Distribution and Energy Distribution |

psr-0010 | EVAP-4, Particle Evaporation from Excited Nuclei |

nesc9952 | EVENT, Explosive Transients in Flow Networks |

nea-0893 | EVGRP, Photo Production MultiGroup Cross-Sections Generated from ENDF/B-4 |

psr-0465 | EVNTRE, Code System for Event Progression Analysis for PRA |

nea-0424 | EXCURS, Heat Transfer Transients in Cylindrical Reactor Channel LOCA |

nea-0228 | EXCURS-3, Reactor Kinetics and Heat Transfer in Cylindrical Channel During Accident |

iaea1273 | EXCURS-3-RR, Kinetics of Research Reactor Reactivity Transient Analysis |

iaea1211 | EXIFON2.0, Neutron, Alpha, Proton, Gamma Emission Spectra |

nesc0321 | EXPALS, Least Square Fit of Linear Combination of Exponential Decay Function |

nea-0311 | EXPANDA-4, 1-D Neutron Diffusion in Slab, Spherical Geometry from JAERI Fast-Set with Criticality Calculation |

nea-0312 | EXPANDA-5, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry for 2 Region Fast Reactor Criticality |

nea-0313 | EXPANDA-6, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry from JAERI Fast-Set with PuO2 UO2 Mixture |

nea-0315 | EXPANDA-70, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry with Criticality Search |

nesc0156 | EXTERMINATOR-2, 2-D MultiGroup Neutron Diffusion in X-Y R-Z or R-Theta Geometry |

psr-0237 | EZVIDEO, DISSPLA Graphics Software Simulation on IBM PC |

psr-0617 | F-SCORE, F-Score Nuclide ID Scoring Applications |

iaea0898 | F5TAB, ENDF/B-4 FILE 5 Data Conversion to Tabulated Form |

nesc9578 | FACET, Radiation View Factor with Shadowing |

ccc-0351 | FALSTF, Neutron Flux and Gamma Flux Detector Response Outside Cylindrical Shields |

nea-0592 | FALT, Orientation of Double Coupled Earthquake Source with Given Amplitudes |

ests0063 | FAMREC, PWR Lateral Mechanical Fuel Rod Assembly Response |

nea-0530 | FANAC, Resonance Parameter by Multilevel Shape Analysis of Neutron Capture Yield Data |

nea-0529 | FANAL, Resonance Parameter by Multilevel Shape Analysis of Neutron Transmission Data |

iaea0868 | FAPCO, Evaluation of Flaws in Nuclear Power Plant Component Structures |

nea-0617 | FAPMAN-IC, LWR Fuel Cost Analysis with Program ORSIM Interface |

nea-0693 | FAPMAN-ORSIM, General Cost Optimization for System of Nuclear Power Plants |

nesc1095 | FASTGRASS, Gaseous Fission Products Release in UO2 Fuel |

psr-0354 | FASTPLOT, Interface Routines to MS FORTRAN Graphics Library |

iaea0835 | FASVER, 2 Group 2-D Diffusion in X-Y Geometry and Adjoint Solution for PWR with Reflector |

nea-0732 | FATAL, General Experiment Fitting Program by Nonlinear Regression Method |

iaea1245 | FDMXPC, ENDF/B Processing, with Reich-Moore and Adler-Adler Resonance Parameter Calculation |

iaea1438 | FE-SHIELDER, shielding thickness of Iron for any photon emitting radionuclide between 0.5 and 10 MeV |

nesc9722 | FE3DGW, Ground Water Flow Model Using Finite Element Method |

psr-0563 | FEAST-METAL-V.1.0, Fuel Engineering and Structural analysis Tool |

nesc1046 | FED, Geometry Input Generator for Program TRUMP |

iaea0830 | FEDGROUP, Group Constant Library from ENDF/B, KEDAK, UKNDL |

ests1121 | FEHM, Finite Element Heat and Mass Transfer Code |

nea-0930 | FELPO, 2-D Minimization of Quadratic Functionals by Finite Elements Method |

nea-0443 | FEM-2D, 2-D MultiGroup Diffusion in X-Y Geometry |

ests0198 | FEM-3, Heavy Gas Dispersion Incompressible Flow |

nea-0545 | FEM-BABEL, 3-D MultiGroup Neutron Diffusion by Galerkin Method |

nea-0566 | FEM-RZ, 2-D MultiGroup Neutron Transport in R-Z Geometry, Eigenvalue and Fixed Source Problems |

nea-1080 | FEMAXI-6, Thermal and Mechanical Behaviour of LWR Fuel Rods |

nea-0478 | FEMB, 2-D Homogeneous Neutron Diffusion in X-Y Geometry with Keff Calculation, Dyadic Fission Matrix |

ccc-0451 | FEMWASTE FEMWATER, Finite Elements Method Waste Transport Through Porous Media |

nesc1144 | FEMWATER BLT, Water or Waste Transport in Soil |

psr-0273 | FERD-PC, Interactive Multichannel Neutron and Gamma Spectrum Matrix Unfolding |

psr-0102 | FERDO/FERD, Unfolding of Pulse-Height Spectrometer Spectra |

psr-0145 | FERRET, Least Square Fit to Nuclear Data and Reactor Physics Problems |

ests0486 | FESH, 2-D Multigroup Neutron Transport with Isotropic Scattering |

ccc-0477 | FEWA-FEMA, Finite Element Method Model of Materials Transport in Ground Water |

nesc9844 | FFSM, Long-Term Nuclear Waste Repository Site Simulation by Monte-Carlo |

nea-1692 | FFT-BM, Code Accuracy Evaluations with the 1D Fast Fourier Transform (FFT) Methodology |

iaea1221 | FIGA, Source Distribution Conversion from X-Y to R-Theta Geometry |

iaea1181 | FINEDAN, Dynamic Stress Analysis in 2-D X-Y and Axisymmetric Geometry |

nea-0896 | FINELM, MultiGroup Diffusion in 3-D by Finite Elements Method |

nea-1901 | FINIX 0.17.12, thermal and mechanical behaviour of a nuclear fuel rod during steady-state and transient conditions |

nea-0310 | FIP-DIG, 1-D Time-Dependent Fission Products Diffusion in Slab, Cylindrical, Spherical Geometry with Gaseous Precursor |

nesc1092 | FIRAC, Nuclear Power Plant Fire Accident Model |

nea-0472 | FIREFLY, X-Ray Diffraction Intensities for Powder Patterns |

nea-0897 | FISP-6, Fission Products Inventory and Energy Release in Irradiated Fuel |

nea-1890 | FISPACT-II 4.0, Inventory Simulation Platform for Nuclear Observables and Materials Science |

nea-0844 | FISPET, MultiGroup Fission Spectra Calculation from ENDF/B |

nea-0706 | FISPIN, Isotope Buildup and Isotope Decay for Actinides, Fission Products, Structure Materials |

nea-0182 | FISPRO-2, Fast Neutron Capture Fission Product Cross-Sections by Hauser-Feshbach with Inelastic Scattering |

nea-0894 | FITOCO, Fine Group to Coarse Group Neutron Flux Conversion for Spectra Unfolding |

iaea1309 | FIXUP2010, ENDF Format Redundant Cross-Sections Check |

uscd1209 | FIZCON, ENDF/B Cross-Sections Redundancy Check |

nesc0395 | FLAC FLAC-SI, Steady-State Flow and Pressure Distribution, 1-D Incompressible Flow Equation |

nea-0636 | FLAKER, Legendre Moments from Scattering Law Tables |

nea-0551 | FLANDES, Flange Design for He Circuits by Taylor-Forge Method |

nesc0689 | FLANGE-ORNL, Flanged Pipe Joint Stress Analysis, Internal Pressure, Moment Loads, Temperature |

nesc0167 | FLARE, 3-D BWR Reactivity and Power Distribution Appraisal Calculation |

nea-0235 | FLARE-JAERI, 3-D BWR and ATR Simulation |

nea-0476 | FLETU, Static Analysis of 3-D Pipeworks by Displacement Method |

nesc9597 | FLODIS, Thermal Response of FSV HTGR Core |

nesc0246 | FLOW-MODEL, Multichannel 2-D 2 Phase Flow for Open Matrix Flow BWR |

nesc9592 | FLOWPLOT2, 2-D, 3-D Fluid Dynamic Plots |

nea-1833 | FLUKA2011.2X.4, Monte Carlo general purpose tool for calculations of particle transport and interactions with matter |

psr-0196 | FLYSPEC, Neutron Spectra Unfolding from Ne213 and Stilbene Scintillation Detectors |

nea-0596 | FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo |

nesc0028 | FOG-1-2-3, 1-D Few-Group Diffusion in Slab Cylindrical Spherical Geometry, Criticality Search, Buckling |

nea-0669 | FONTA, Radiation Release in Atmosphere and Deposition in Human Organs |

nesc0174 | FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients |

psr-0092 | FORIST, Ne-213 Scintillation Detector Neutron Spectra Unfolding |

nea-0810 | FORM-OTA, MultiGroup Constant for Epithermal Neutron Slowing-Down in Homogeneous Media |

nesc0514 | FORSIM, Solution of Ordinary or Partial Differential Equation with Initial Conditions |

psr-0078 | FORSIM-6, Automatic Solution of Coupled Differential Equation System |

iaea1388 | FOTELP-2014, Photons, Electrons and Positrons Transport in 3D by Monte Carlo Techniques |

nea-0867 | FOURACES, MultiGroup Cross-Sections, Resonance Calculation from ENDF/B, KEDAK, UKNDL |

nea-0593 | FPSPH DFPSPF, Line Shape Function for Doppler Broadened Resonance Cross-Sections Calculation |

ccc-0603 | FPZD, Reactor Burnup by MultiGroup Neutron Diffusion |

nesc9411 | FRACFLO, 2-D Radionuclide Groundwater Transport in Fracture System |

nea-0465 | FRAMES, Vibration Analysis of Spaceframes with Lumped Mass Distribution |

nesc9915 | FRAMIS, Relational Data Base Management System |

nea-0396 | FRANCESCA, 2 Phase Flow Dynamic in Boiling Test Channel and Heat Elements Conduction |

nea-0397 | FRANCESCA-BWR, 2 Phase Flow Dynamic for BWR Cooling Channel |

psr-0363 | FRANCO, Finite Element Method (FEM) Fuel Rod Analysis for Solid and Annular Configurations |

nesc0694 | FRAP-S3 FRAP-S1, Steady-State LWR Oxide Fuel Elements Behaviour |

nesc0658 | FRAP-T, Temperature and Pressure in Oxide Fuel During LWR LOCA |

nesc0694 | FRAPCON2, Steady-State LWR Oxide Fuel Elements Behaviour, Fission Products Gas Release, Error Analysis |

nesc0479 | FREADM-1, Reactor Kinetics Thermohydraulics Calculation for Fast Reactor Accidents |

nea-0692 | FRELIB, Failure Reliability Index Calculation |

nea-0982 | FRETA-B, LWR Fuel Rod Bundle Behaviour During LOCA |

nesc0301 | FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements |

nesc9659 | FRTGEN, Fault Trees by Subtree Generator from Parent Tree for Program FTAP |

nesc9659 | FRTPLT, Fault Tree Structure and Logical Gates Plot for Program FTAP |

nea-1846 | FSKY4C, Gamma Ray Skyshine Analysis Code |

nesc0666 | FTA, Fault Tree Analysis for Minimal Cut Sets, Graphics for CALCOMP |

nesc9659 | FTAP, Minimal Cut Sets of Arbitrary Fault Trees |

nesc9860 | FTRANS, Radionuclide Flow in Groundwater and Fractured Rock |

nea-1812 | FUELPERFORMANCE-REP, Seminars on nuclear fuel performance based on basic underlining phenomena, proceedings |

iaea1303 | FUP1, Fast Neutron Cross-Sections for Fissile Nuclei by Hauser-Feshbach Theory |

nea-1021 | FURNACE, Neutronic Calculation in 3-D Toroidal Geometry |

nea-0314 | FURNACE-J, 2-D Diffusion Burnup for Fast Reactors from JAERI Fast-Set |

nesc0862 | FX2-TH, 2-D MultiGroup Neutron Diffusion in X-Y, R-Z and R-Theta Geometry with Thermal Feedback |

ccc-0494 | G33-GP, Multigroup Gamma Scattering Using Geometric Progression Buildup Factors |

nesc0223 | GAD-2, Fuel Cycle Depletion Calculation with Partial Refueling and Fuel Recycling |

psr-0610 | GADRAS-DRF-18.7.6, Gamma Detector Response and Analysis Software-Detector Response Function |

nea-0005 | GAKER-KIRA, Energy Transfer of Protons in H2O or Polyethylene and Deuterons in D2O |

nesc0310 | GAKIN-2, 1-D MultiGroup Time-Dependent Neutron Diffusion, Finite Difference Method |

nea-1459 | GALIST, Decay Gamma Spectra Retrieval from ENSDF |

nesc0033 | GAM, Slowing-Down Neutron Spectra in P1 and B1 Approximation, MultiGroup Constant |

nesc9654 | GAMANAL, Radioactive Species Mixtures by Gamma Spectra Analysis |

nea-1175 | GAMFIL, Photon Production Cross-Sections in ENDF/B Format |

psr-0154 | GAMIDENT, Aid Identification of Unknown Materials by Gamma-Ray Spectroscopy |

ccc-0042 | GAMLEG-JR, MultiGroup Gamma Cross-Sections, Energy Absorption Coefficient Generator for Transport Calculation |

nea-0268 | GAMMONE, Multi-Region Shield Gamma Penetration from Various Geometries Source by Monte-Carlo |

nea-1827 | GANAPOL-ABNTT, Analytical Benchmarks for Nuclear Engineering Applications, Case Studies in Neutron Transport Theory |

nea-1852 | GANDR/SEMOVE, Program for Calculating Derivatives of Processed Multigroup Nuclear Data by Discrete Differences |

nesc0770 | GAPCON-THERMAL3, Fuel Rod Steady-State and Transient Thermal Behaviour, Stress Analysis |

nesc0606 | GAPER-1D, 1-D MultiGroup 1st Order Perturbation Transport Theory for Reactivity Coefficient |

nesc0317 | GAPOTKIN, Space-Independent Reactor Kinetics for a General Reactivity Function |

nea-1601 | GARDEC, Estimation of dose-rates reduction by garden decontamination |

nesc0263 | GASKET-2, Thermal Neutron Scattering Law for Moderators, Harmonic Vibrations and Gaseous |

iaea0877 | GASPAN-ZKD, Ge(Li) Detector and Multichannel Analyser Gamma Spectra Unfolding |

ccc-0463 | GASPAR-II, Radiation Exposure to Man from Air Releases of Reactor Effluents |

nesc0380 | GATT, 3-D Few-Group Neutron Diffusion for Power Distribution in Hexagonal Reactor Core for HTGR |

nesc0622 | GAUSS-6, Experimental Gamma Spectra Analysis, Isotope Identification, Decay Rates |

nesc0232 | GAZELLE-5, Gas Cooled Fast Reactor Core Design and Core Performance |

iaea1362 | GCASCAD, Gamma Production Cross Sections Statistical Model |

nea-1864 | GEF 2018/1.1, Code for Simulation of Nuclear Fission Process |

nesc0576 | GEM, Fuel Cycle Cost and Economics for Thermal Reactor, Present Worth Analysis |

nea-1652 | GEM, Monte-Carlo Code for Simulating a Decaying Process of an Excited Nucleus |

ests0742 | GENAEA, Alpha Spectra Unfolding |

nea-0606 | GENDY, Reactor Dynamic Program with Variable Time Step Control |

ccc-0737 | GENII 2.10, Environmental Radiation Dosimetry System |

ccc-0601 | GENII-LIN, Multipurpose Health Physics Code |

nea-0605 | GENP-2, Program System for Integral Reactor Perturbation |

uscd1210 | GETMAT, ENDF/B Material Retrieval |

nesc0887 | GETOUT, Radioactive Release and Decay Chain Calculation for Nuclear Waste Disposal |

nea-0584 | GFX/GAMP1, Above-Ground Radiation Field from Terrestrial K, U, Th Gamma Emitters |

nea-0543 | GGTC-ENEL, MultiGroup Neutron Spectra in P1, B1, B2, B3 Approximation and Thermos Calculation |

nea-0073 | GHT, 3-D Steady-State and Transient Heat Conduction |

psr-0229 | GIP, Group Organized Cross-Sections Library for ANISN, DOT |

psr-0304 | GIRAFFE, Isotope Release Analysis in LMFBR Fuel Elements Failure |

psr-0192 | GLUCS, Experimental Reaction Cross-Sections Evaluation for ENDF/B-5 |

psr-0367 | GMA, Generalized Least-Squares Cross-Sections Evaluation for ENDF Format |

psr-0125 | GNASH-FKK, FKK, Preequilibrium, Statistical Model Cross-Sections and Emission Spectra |

iaea1271 | GNOMER, Core Power Distribution by 1-D, 2-D, 3-D MultiGroup Neutron Diffusion |

nea-0535 | GOLIA-RK, Structure Stress for Isotropic Materials with Creep and Temperature Fields |

nea-0550 | GOMESH, Finite Elements Structure Plot with Triangular Mesh |

nesc0046 | GRACE-2, MultiGroup Multi-Region Gamma Attenuation Gamma Dose in Cylindrical or Spherical Geometry |

uscd1211 | GRALIB, DISSPLA Plot Routines Emulator |

iaea1175 | GRAP, Gamma-Ray Level-Scheme Assignment |

nea-1043 | GRAPE, System for Precompound and Compound Nuclear Reactions |

nesc0624 | GRAPH, Data Processing, Statistical Analysis, Correlations and Graphics |

ests0075 | GRASS-SST, Fission Products Gas Release and Fuel Swelling in Steady-State and Transients |

nesc9911 | GRAY CNVUFAC, Black-Body Radiation View Factors with Self-Shadowing |

iaea0908 | GRENADE, Green's Function Nodal Algorithm for Diffusion Equation |

psr-0231 | GRESS-3.0, FORTRAN Precompiler with Differentiation Enhancement |

nea-0433 | GRETEL, Ge(Li) Gamma Spectra Unfolding |

ests0576 | GRIDMAKER, 2-D, 3-D Finite Element Method Grid Generation for Ground Water and Pollutant Transport |

nesc0620 | GROUP-2, Atomic and Molecular Lattice Vibrations, Group Theory and Symmetry |

iaea0849 | GROUPIE2010, Bondarenko Self-Shielded Cross Sections from ENDF/B |

nea-1111 | GROUPXS, MF6 Format ENDFB-6 Continuum Region Diffusion Cross-Sections Processing |

psr-0321 | GRPANL, Ge Gamma and Alpha Detector Spectra Unfolding |

ccc-0774 | GRSAC, Graphite Reactor Severe Accident Code |

nea-1690 | GRTUNCL-3D/R-THETA-Z, Code to Calculate Semi Analytic First Collision Source and Uncollided Flux in an R-theta-Z grid |

ccc-0721 | GRTUNCL3D, Code to Calculate Semi Analytic First Collision Source and Uncollided Flux (X, Y, Z) |

nea-1899 | GRUCON-D-2017-01, Data Processing for Evaluated Working libraries (transport and shielding) |

ccc-0276 | GRUNCLE, 1st Collision Source Calculation for Program DOT |

nesc9845 | GSM, Columbia-Plateau Geologic Repository Site Long-Term Evolution Simulation |

nea-1400 | GTM-1, Radionuclide Transport Through Ground Water |

nesc0618 | GTR2 GAPCON-THERMAL2, Steady-State Fuel Rod Thermal Behaviour and Fission Products Gas Release |

nea-1820 | GTSP, automatic ultrasonic inspection of Guide Tube Support Pin in nuclear power plants |

ccc-0697 | GUI2QAD, Graphical Interface for QAD-CGPIC, Point Kernel for Shielding Calculations |

nea-0876 | H2O, Calculation of Thermodynamics Properties of Steam and H2O |

nea-0682 | H2OTP, Temperature Dependent and Pressure Dependent Thermodynamics Properties, Transport Properties of H2O |

nesc0443 | HAA3B, Heterogeneous Aerosol Transport after LMFBR Accidents, Lognormal Size Distribution |

nesc0797 | HAARM, Time-Dependent Diffusion and Deposition of Radioactive Aerosols, LMFBR Accidents |

ccc-0665 | HABIT 1.1, Toxic and radioactive release hazards in reactor control room |

ests1100 | HABIT, Toxic and Radioactive Release Hazards in Reactor Control Room |

ccc-0452 | HADOC, External and Internal Organ Doses from Radiation Release at Hanford |

iaea1222 | HAMCIND, Cell Burnup with Fission Products Poisoning |

nesc0277 | HAMMER, 1-D MultiGroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation |

ccc-0387 | HARAD, Decay Isotope Concentration from Atmospheric Noble-Gas Release |

nea-1345 | HARPHRQ, Geochemical Reaction Modelling |

nea-0547 | HASSAN, Time-Dependent Temperature Distribution and Stress and Strain in HTR Fuel Pins |

nesc0830 | HAUSER-5, Capture and Fission Cross-Sections Using Hauser-Feshbach with Woods-Saxon Potential |

nesc9819 | HCT, Time Dependent 1-D Gas Hydrodynamics, Chemical Kinetics, Chemical Transport |

iaea1330 | HEATER, Reaction Rate Tables from Cross-Sections with Weighting |

nea-1292 | HEATHYD, Steady-State Thermal Hydraulic Analysis of Low-Enriched U Fuel Reactor |

psr-0199 | HEATING-7, Multidimensional Finite-Difference Heat Conduction Analysis |

nea-1095 | HEATP, Steady-State and Transient Heat Transfer in PWR |

nea-0303 | HEATRAN, 2-D Heat Diffusion for X-Y or R-Z Geometry with Heat Transfer Across Gaps |

nea-0490 | HEDO-2, Magnetic Field Calculation and Plot of Air Core Coils |

nea-0302 | HEITLER, Compton Cross-Sections, Photoelectric Cross-Sections, Pair-Production Cross-Sections, Total Cross-Sections |

nesc0775 | HEMP, 2-D Elastic Plastic Flow in 2-D X-Y or Cylindrical Geometry by Lagrangian Method |

nea-1666 | HEPROW, Unfolding of pulse height spectra using Bayes theorem and maximum entropy method |

nea-0536 | HERA-1A, Steady-State Thermohydraulics of Na Cooled Fuel Rod Bundles |

nesc0136 | HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor |

nesc0527 | HERMES, Regional Release of Radionuclides from Reactor Plant Operation |

nea-1265 | HERMES-KFA, High-Energy Radiation Transport by Monte-Carlo |

nea-0176 | HETERO, Flux and Power Distribution in Thermal Reactor by 3-D, 2 Group Line Source Sink Method |

iaea1240 | HEXAB-3D, 3-D Few-Group Diffusion for Hexagonal Core Geometry |

nea-1125 | HEXANN-EVALU, Neutron Irradiation of Reactor Pressure Vessels |

nea-0481 | HEXCO-H, Coherent Elastic Scattering and Inelastic Scattering in Hexagonal Isotropic Crystal |

iaea0914 | HEXNOD23, 2-D, 3-D Coarse Mesh Solution of Steady State Diffusion Equation in Hexagonal Geometry |

iaea1317 | HFMOD, Elastic and Inelastic Cross-Sections Calculation by Hauser-Feshbach and Moldauer |

iaea0954 | HFTT, Nuclear Reaction Cross-Sections by Compound-Nucleus Evaporation Model |

ests0545 | HGSYSTEM, Atmospheric Dispersion for Ideal Gases and Hydrogen Fluoride (HF) |

ests1242 | HGSYSTEMUF6, Simulating Dispersion Due to Atmospheric Release of Uranium Hexafluoride (UF6) |

iaea1253 | HOMO, Homogenization of ANISN and DOT Condensed Cross-Sections Output |

nea-1169 | HORN, Fission Products Transport in Primary Coolant System of BWR and PWR in LOCA |

ccc-0644 | HOTSPOT 3.0.1, Health Physics Code System for Evaluating Accidents Involving Radioactive Materials |

ests0648 | HTRATE, Power Plant Heat Rate Improvement from Condenser Retubing |

nea-0518 | HUBBLE-BUBBLE, Transient Subcooled or Superheated H2O Bubble Flow |

iaea1377 | HYDMN, Thermal Hydraulics of Miniature Neutron Source Reactor |

nesc9553 | HYDRA-2, 3-D Heat Transport for Spent Fuel Storage System |

nea-0499 | HYDY-B1, Channel Thermohydraulics During LOCA of BWR, PWR |

ests0405 | HYFRAC3D, 3-D Hydraulic Rock Fracture Propagation by Finite Element Method |

ests0406 | HYFRACP3D, 3-D Hydraulic Fracture Propagation by Finite Element Method |

psr-0101 | HYPERMET, Ge(Li) Detector Multichannel Analyser Gamma Spectra Evaluation |

nea-0100 | HYTHEST, Dependence of Fuel Fabrication Tolerances on Hydraulics of BWR, PWR |

nea-0216 | HYTRAN, Open Channel Thermal and Hydraulic Transients in LOCA |

nea-0995 | IBIS, FBR 3-D Steady-State and Kinetics with Thermohydraulic Feedback |

iaea0974 | ICAR, Nuclear Level Density by Free-Gas or BCS Nuclear Models |

nea-0329 | ICAROG, WIMS-D/4 Library Utility |

nesc9683 | ICARUS-LLNL, 1-D Heat Transfer in Planar, Cylindrical, Spherical Geometry Using Finite Element Method |

ests0167 | ICCG2, 2-D Partial Differential Equations Linear Symmetric Matrix Solver |

ests0168 | ICCG3, 3-D Partial Differential Equations Linear Symmetric Matrix Solver |

ccc-0651 | ICOM, Ion Radiation Transport Calculation for Shielding and Dosimetry |

nea-0353 | ICON, Reactor Operation Fission Products Inventory Calculation |

nea-1823 | ICRS1, Proceedings of the First Radiation Shielding Symposium, Cambridge, UK 1958 |

nea-1326 | IFF, Full-Screen Input Menu Generator for FORTRAN Program |

ests0169 | ILUCG2, 2-D Partial Differential Equations Asymmetric Matrix Solver |

ests0170 | ILUCG3, 3-D Partial Differential Equations Linear Asymmetric Matrix Solver |

ests0005 | IMPACTS-BRC2.1, General Radiological Impacts Analysis |

nesc0779 | IMPORTANCE, Minimal Cut Sets and System Availability from Fault Tree Analysis |

iaea1378 | INDOSE V2.1.1, Internal Dosimetry Code Using Biokinetics Models |

nea-0485 | INDRA, Fusion Reactor Blanket Neutronics, Gamma Heating, H3 Breeding |

nesc0609 | INDX, X-Ray Diffraction Powder Pattern Indexing, Trial Unit Cell Testing |

iaea1248 | INDXENDF, Preparation of Visual Catalogue of ENDF Format Data |

psr-0313 | INFLTB, Dosimetric Mass Energy Transfer and Absorption Coefficient |

nesc0975 | INGEN, 2-D, 3-D Mesh Generator for Finite Elements Program |

ccc-0185 | INREM-EXREM-3, Time-Dependent Organ Doses from Isotope Inhalation and Ingestion |

nea-0744 | INTEGR, Escape Transmission Probability in 1-D Cylindrical Geometry for Program FACEL |

uscd1212 | INTER, ENDF/B Thermal Cross-Sections, Resonance Integrals, G-Factors Calculation |

iaea0886 | INTERTRAN-I and INTERTRAN-II, Radiation Exposure from Vehicle Transport of Radioactive Material |

uscd1213 | INTLIB-6, Graphic Device Interface Library for ENDF/B Processing Codes |

psr-0054 | INTRIGUE-2L, Subroutines for Linear, Log, Semi-Log CALCOMP Plotter |

nea-1154 | INTRUDE, Radiation Risk from Intrusion into Shallow Land Waste Storage Site |

nea-1153 | INVENT, Dose Rates, Inhalation, Ingestion Risk from Closed Waste Storage Site |

nea-1894 | INVENTDYN.MT, calculates the dynamics of the amount of isotope and its daughter nuclides with time stamps |

ccc-0365 | IODES, Calculating the Estimation of Dose to the World Population from Releases of Iodine-129 to the Environment. |

ccc-0526 | IONMIG, Radionuclide Migration Through Porous Media |

iaea0901 | IPEET-103, Neutron Induced Reaction Cross-Sections for Fissile Nuclides, Preequilibrium Model |

nea-1821 | IPLOT, interactive MELCOR data plotting system |

ests0109 | IRDAM, Interactive Rapid Dose Assessment from Reactor Accident Effluents |

nea-0513 | IRESINT-3, Resonance Absorption in Square or Hexagonal Lattice by Single-Level Breit-Wigner |

ests0003 | IRRAS, Integrated Reliability and Risk Analysis System for PC |

iaea1328 | ISABEL EVA PACE-2, Evaporation Model with Intranuclear Cascade Input |

ccc-0636 | ISO-PC, X-Ray, Gamma-Bremsstrahlung Dose-Rates |

ccc-0079 | ISOSHLD, Decay Gamma Dose, Bremsstrahlung Dose Behind Shield, Fission Products Source Strength |

nea-0434 | ISOTEX-1, Time-Dependent Heavy Isotope and Fission Products Concentration in U Reactor or Pu Reactor |

iaea1229 | ISOTHERM, Ion-Exchange IsoThermal Calculation and Plot |

ests0219 | ITOUGH2, Inverse Modeling for TOUGH2 Multiphase Flow Simulators |

ccc-0467 | ITS, TIGER System of Coupled Electron Photon Transport by Monte-Carlo |

iaea0940 | JADSPE, Multi-Channel Gamma Spectra Unfolding Program |

nesc1058 | JAKEF, Gradient or Jacobian Function from Objective Function or Vector Function |

nea-1760 | JANIS 4.0, a Java-based nuclear data display program |

nea-1838 | JASMINE V.3, Steam explosion simulation |

nea-1811 | JDL-IMPORTANCE, Adjoint Function: Physical Basis of Variational & Perturbation Theory in Transport & Diffusion Problems |

nea-1843 | JDL-REACTOR-KINETICS, Nuclear Reactor Kinetics and Control |

nea-1844 | JDL-THERMODYNAMICS, Thermodynamics: Frontiers and Foundations |

nea-0317 | JFUSER, JAERI Fast-Set Group Constant Collapsing and Data Conversion for Program LTFR-70 |

nea-1871 | JN-METD, N Transport with Isotropic Scattering, Bare Slabs and Homogeneous Slabs (JN-METD1), Multilayer Slabs (JN-METD2) |

psr-0008 | JOMREAD, Check of 3-D Geometry Structure from Quadratic Surfaces |

nea-0624 | JOSHUA, Neutronics, Hydraulics, Burnup, Refuelling of LWR |

nea-0154 | JPHYDRO, Voids and Flow Velocity in Steady-State BWR System |

nesc0877 | K-FIX(3D), Transient 2 Phase Flow Hydrodynamic, X-Y-Z and Cylindrical Geometry, Eulerian Method |

nesc0727 | K-FIX, Transient 2 Phase Flow Hydrodynamic in 2-D Planar or Cylindrical Geometry, Eulerian Method |

nesc0876 | K-TIF, Thermohydraulic Dynamic of PWR in Steady-State and Transient Flow Conditions |

nea-0492 | KAMCCO, 3-D Time-Dependent Homogeneous and Inhomogeneous Neutron Transport by Monte-Carlo Method |

psr-0306 | KAOS-V, Neutron Fluence to Kerma Factor Evaluation from ENDF/B-5 and JENDL-2 |

nea-0343 | KASY, 3-D Homogeneous Neutron Diffusion in X-Y-Z, R-Theta, Hexagonal-Z Geometry by Synthesis Method |

nea-1824 | KCUT, code to generate minimal cut sets for fault trees |

nesc0556 | KEELE, Minimization of Nonlinear Function with Linear Constraints, Variable Metric Method |

nea-0578 | KEMA, KEDAK Utility, Data Update |

ccc-0510 | KENO-4(RG), KENO-4 with Random Geometry |

ccc-0436 | KENO-4/CRC, MultiGroup 3-D P1 Scattering Monte-Carlo Transport Calculation with Neutron Balance Edit |

nea-1467 | KENO-VA-PVM KENO-VA-SM, KENO5A for Parallel Processors |

psr-0541 | KENO2MCNP, Version 5L, Conversion of Input Data between KENOV.a and MCNP File Formats |

psr-0450 | KENO3D, Visualisation Tool for KENO V.A and KENO-VI Geometry Models |

ccc-0548 | KENO5A-PC, Monte-Carlo Criticality with Supergrouping |

nea-0288 | KERBREK, Fuel Cycle Cost Analysis for Power Reactor |

nea-1865 | KICHE 1.3, Kinetics of Iodine Chemistry in the Containment of LWRs under Severe Accident Conditions |

nea-0616 | KIM, Steady-State Transport for Fixed Source in 2-D Thermal Reactor by Monte-Carlo |

nea-0112 | KINAX-3, 1-D 1 Group Reactor Kinetics with Xe and I and Fission Products Heating and Auto-Control |

nea-1002 | KINE, 1-D PWR Dynamic with Partial Core Boiling |

iaea1339 | KINETIC, Time-Dependent Heat and Mass Transfer |

nea-1293 | KINIK, Absorber Rod Calibration Kinetics |

nesc0528 | KITT, Component and System Reliability Information from Kinetic Fault Tree Theory |

ests0154 | KIVA3, Transient Multicomponent 2-D and 3-D Reactive Flows with Fuel Sprays |

nea-1001 | KORIGEN, Isotope Inventory, Radiation Heat from PWR Burnup |

nea-0417 | KOSAK, Power Plant Cost Optimization with Pu Availability Option |

nea-0441 | KPD, Time-Dependent Fuel Cycle Cost Calculation for Various Reactor Types |

ccc-0229 | KRONIC, Annual Body Tissue Dose from Continuous Atmospheric Release |

nesc9520 | KRYSI, Ordinary Differential Equations Solver with Sdirk Krylov Method |

nea-0342 | KTOE, KEDAK to ENDF/B Format Conversion with Linear Linear Interpolation |

nesc0987 | L2RMAT, L**2 Method of R Matrix Propagation |

iaea1232 | LABAN-PEL, 2-D MultiGroup Neutron Diffusion in X-Y Geometry by Response Matrix Eigenvalue |

nesc0992 | LADTAP-2, Organ Doses to Man and Other Biota from Aquatic Environment |

ccc-0696 | LAHET 2.8, Code System for High Energy Particle Transport Calculations |

psr-0020 | LAPHAN0, P0 Gamma Production Matrices from ENDF/B |

nesc0249 | LASER, Slowing-Down Neutron Spectra and Burnup for Thermal Reactors, Neutron Transport Theory |

nea-0573 | LASER-PNC, Neutron Spectra in Uniform Lattice with Burnup Calculation |

nesc0691 | LASIP-3, CCCC Utility for BCD to BIN Conversion and BIN Data Listing |

nesc0918 | LASO, Subroutine Library for Matrix Manipulation, Eigenvalues and Eigenvectors |

nea-0192 | LAZY, General Experimental Data Processing Program |

ests0463 | LDEF-SS, Solve Equation Two Phase Fluid Flow in Spray Dryers |

nea-0479 | LEAP, Scattering Law for Continuous Phonon Spectra |

iaea1310 | LEGEND2010, Angular Distribution Table Calculations in ENDF Format |

nesc0279 | LEOPARD, Fast and Thermal Neutron Spectra from Temperature and Geometry with Depletion Calculation |

ccc-0343 | LEOPARD-MICRO, Spectrum-Dependent Non-Spatial Fuel Depletion |

psr-0277 | LEPRICON, PWR Vessel Dose Analysis with DORT and ANISN Program |

nesc9426 | LFK, FORTRAN Application Performance Test |

nea-0124 | LGH, Gamma Streaming and Neutron Streaming for Duct |

psr-0394 | LHS, Multivariate Sample Generator by Latin Hypercube Sampling |

nesc1085 | LHS-ESTSC, Multivariate Sample Generator by Latin Hypercube Sampling |

iaea0902 | LIANG, Neutron Induced Compound Nucleus Reaction Cross-Sections by Statistical Model |

nea-0167 | LIE-PN, Pn Neutron Transport in Radial Geometry Cell with Source Problems Calculation |

nea-1337 | LIMES, IMF in Heavy Ion Nuclear Reaction by Sum-Rule Model |

nesc0657 | LINDA, Diagnostics of Stress Analysis of Linear Elastic Structure by Least Square Fit |

iaea1311 | LINEAR2010, Linear-Linear Interpolation of ENDF Format Cross-Sections |

nesc0800 | LINPACK, Subroutine Library for Linear Equation System Solution and Matrix Calculation |

iaea1331 | LINTAB, Linear Interpolable Tables from any Continuous Variable Function |

psr-0117 | LINX, MINX Library Utility, Data Merge |

nea-0860 | LISA, Hazard Assessment of Nuclear Waste Disposal in Geological Formations |

uscd1214 | LISTEF, ENDF/B Data File Summary List |

nesc0638 | LISTF-4, ENDF/B Utility, Data Listing |

nea-0623 | LOCA-MARK-2, Fuel Temperature and Clad Temperature in HWR Steam Generator LOCA |

nea-0965 | LOLA-SYSTEM, JEN-UPM PWR Fuel Management System Burnup Code System |

nea-0185 | LOOP-3, Hydraulic Stability in Heated Parallel Channels |

nea-1026 | LOUHI, Generator Spectra Unfolding Program with Linear and Nonlinear Regularization |

iaea1304 | LPA1, LPA2, Deconvolution Program Using Fourier Transform |

nesc9449 | LPGC, Levelized Steam Electric Power Generator Cost |

ccc-0385 | LPGS, Radiation Exposure from Radioactive Release into Hydrosphere |

ccc-0064 | LPSC, Protons and Neutron Flux, Spectra Behind Slab Shield from Protons Irradiation |

iaea1260 | LPTAU, Quasi Random Sequence Generator |

nesc9721 | LRSYS, PASCAL LR(1) Parser Generator System |

nesc1033 | LSAP-DIGLIB, Linear Control System Design, Analysis, Plotting |

nea-1306 | LSHINSE, Air Scattering Neutron and Gamma Doserates for Complex Shielding Geometry |

psr-0233 | LSL-M2, Neutron Spectra Log Adjustment for Dosimetry Applications |

uscd1227 | LSODA, Ordinary Differential Equation Solver for Stiff or Non-Stiff System |

uscd1228 | LSODAR, Ordinary Differential Equation Solver for Stiff or Non-Stiff System with rootfinding |

uscd1223 | LSODE, 1st Order Stiff or Non-Stiff Ordinary Differential Equations System Initial Value Problems |

uscd1229 | LSODES, Ordinary Differential Equations System Sparse Matrices |

uscd1224 | LSODI, Implicit Ordinary Differential Equations System Either Dense or Banded Matrices |

uscd1225 | LSODIS, Implicit Ordinary Differential Equations System Sparse Matrices |

ests0264 | LSODKR, Stiff Ordinary Differential Equations (ODE) System Solver with Krylov Iteration and Rootfinding |

uscd1230 | LSODKR, Stiff Ordinary Differential Equations (ODE) System Solver with Krylov Iteration with Rootfinding |

uscd1231 | LSODPK, Ordinary Differential Equations Solver for Stiff and Nonstiff System with Krylov Corrector Iteration |

uscd1226 | LSOIBT, Implicit Ordinary Differential Equations System Block Tridiagonal Matrices |

iaea1268 | LSQXY, Curve Fitting with Uncertainty Weighting |

nea-0316 | LTFR-4, Library Generated for Fast Reactor Design Program from JAERI Fast-Set MultiGroup Constant |

nesc0648 | LUGS, Stress Analysis, Flexibility Factors for Rectangular Attachment on Thin Shell |

ccc-0220 | LUIN-II, Cosmic Ray Cascade Generator and Particle Fluxes |

nea-0250 | LUPO, Temperature and Void Rate and Pressure Drop and Flow Rate in Pressure Loop |

ccc-0631 | LWRARC, PWR and BWR Spent Fuel Decay Heat Generator |

nesc0381 | LYNNE, Inelastic Scattering by Multipole Expansion of Woods-Saxon |

psr-0132 | MACK, Fluence to Kerma Generator from ENDF/B |

nesc0574 | MACS, Lattice Vibrations Structure Factors for Thermal Neutron Scattering in Moderators |

nea-0836 | MADONNA, Neutron Flux with Void Region by Removal Diffusion Method |

nesc1006 | MAEROS, Multicomponent Aerosol Time Evolution |

ccc-0359 | MAGIK, Photon Dose Rates from Nucleon-Nucleus % Meson-Nucleus Collisions |

ests0386 | MAGNUM-2D, Heat Transport and Groundwater Flow in Fractured Porous Media |

nea-0931 | MAIA, Eigenvalues for MHD Equation of Tokamak Plasma Stability Problems |

nea-0565 | MAILLE, Triangular Finite Elements Generator for Planar Structure |

nesc0256 | MANTA, Heat Transfer Fuel Elements Cluster to Single-Phase Steady-State Fluid Flow |

nea-1047 | MANYCASK, Radiation Dose Rate Around Many Casks |

nea-1096 | MAPLE, Fault Tree Plotting |

nea-0517 | MAPLIB, Thermodynamics Materials Property Generator for FORTRAN Program |

nesc0939 | MAPPER, Graphics for Transparencies and Slides Using DISSPLA System |

nea-0528 | MARC, MultiGroup Diffusion in R-Z, X-Y, R-Theta, Slab, Cylindrical, Spherical, Hexagonal, Triangular Geometry |

nesc0734 | MARCH, Containment Behaviour after LOCA, Blowdown, Meltdown, Metal H2O Reaction |

nea-1017 | MARCOPOLO, Radial and Axial Diffusion Coefficient for Cylindrical Wigner-Seitz Reactor Cell |

nea-0526 | MARE, Reaction Cross-Sections by Blatt-Ewing Statistical Evaporation Model |

nea-0926 | MARIA-SYSTEM, PWR Fuel Assembly Calculation for Program CARMEN-System and Simulation |

ccc-0503 | MARINRAD, Health Hazard from Radioactive Material Release into Ocean |

psr-0137 | MARLOWE 15b, Computer Simulation of Atomic Collisions in Crystalline Solids |

nea-1307 | MARMER, Point-Kernel Shielding Calculation with Nuclide Concentrations from ORIGEN-S |

psr-0117 | MARS-ORNL, Processing Program Collection for AMPX, CCCC, ANISN, DOT, MORSE Format Library |

nea-0983 | MARTHA, Nai(Tl) Gamma Scintillation Detector Response by Monte-Carlo |

ests0212 | MASCON, Mass-Consistent Atmospheric Flux Model |

nesc9522 | MASCOT, Multi Dim Groundwater Transport of Radioactive Waste |

nesc0745 | MATADOR, Fission Products Release and Deposition in LWR Containment, Meltdown Accident |

ests0279 | MATHEW/ADPIC, Air Concentration and Ground Deposition from Point Sources |

nea-0380 | MATRA, Void Simulation in Steam and H2O Mixture Channel in Accident |

uscd1159 | MATXTST, Basic Operations for Covariance Matrices |

psr-0130 | MATXUF, Ne-213 Liquid Scintillation Detector Neutron Spectra Unfolding |

psr-0001 | MAX-XTREME, 1 Constraint Lagrange Multipliers for 25 Variables |

ests0221 | MAXWELL3, 3-D FEM Electromagnetics |

nesc0355 | MC**2-2, MultiGroup Neutron Spectra, Slowing-Down Calculation Using ENDF/B, P1 and B1 Approximation |

psr-0350 | MC*2-2, Calculation of Fast Neutron Spectra and Multigroup Cross-Sections from ENDF/B Data |

uscd1241 | MCART, solve the time dependent neutron transport equation |

nea-1643 | MCB1C, Monte-Carlo Continuous Energy Burnup Code |

nea-0452 | MCDATA, MC**2 Cross-Sections Conversion for Programs CITATION, ANISN, DOT |

nea-1632 | MCDSIM, Atmospheric Monte Carlo Dispersion Simulation |

nea-1733 | MCNP4B-GN, Monte Carlo Code System for (gamma,n) production and transport in high-Z materials |

iaea0889 | MCRAC, In Core Fuel Management, Program of PFMP System |

nea-0971 | MCRTOF, Multiple Scattering of Resonance Region Neutron in Time of Flight Experiments |

ests1678 | MCSLTT, Monte Carlo Simulation of Light Transport in Tissue |

nea-1859 | MCUNED, MCNPX Extension for Using light Ion Evaluated Nuclear Data library |

nea-1166 | MCVIEW, 3-D Radiation View Factor by Monte-Carlo Method |

ccc-0156 | MECC-7, Medium-Energy Intranuclear Cascade Code System |

nea-0362 | MEDEA, Steady-State Pressure and Temperature Distribution in He H2O Steam Generator |

nea-1140 | MEDUSA-1B, 1-D Plasma Hydrodynamic Analysis of Fusion Pellet Ion Beams |

nea-0583 | MEDUSA-PIJ, 1-D Thermohydraulic Analysis of Laser Driven Plasma |

nea-1889 | MEGA, Mechanistic and Engineering Fission Gas Release Model for a Uranium Dioxide Fuel |

nea-1057 | MELODIE, Radiological Assessment of Nuclear Waste Migration in Ground Water |

nesc0700 | MELT-3, Thermohydraulics and Neutronics, Fast Reactor Transients with Feedback |

nea-0351 | MERCURE, 3-D Gamma Heating and Gamma Dose Rate and Fast Flux by Monte-Carlo |

nea-0194 | MERCURE-3, Gamma Attenuation by Line-of-Flight in 3-D Heterogeneous Geometry |

iaea1312 | MERGER2010, Merges ENDF/B Data by Material Number or Identifier |

nesc0825 | MESA, Fourier Analysis of Maximum Entropy Spectra and Correlation Function Analysis |

nea-0346 | MESHGEN, Triangular Finite Elements Generator |

nea-0348 | MESHPLOT, CALCOMP Plot of 2-D Triangular Finite Elements Mesh |

nea-0347 | MESHREF, Finite Elements Mesh Combination with Renumbering |

nesc9862 | MESOI2.0, Atmospheric Transport of Effluent Puffs |

ests0331 | MESORAD, Emergency Response Airborne Dose Assessment |

nea-1534 | MESYST, Simulation of 3-D Tracer Dispersion in Atmosphere |

iaea1387 | MEXP, EXTERMINATOR-2 Utility Programs |

nesc9479 | MGA, Pu Isotope Abundance from Multichannel Analyzer Gamma Spectra |

psr-0542 | MGA8, Determine Pu Isotope Abundances from Multichannel Analyzer Gamma Spectra |

ests0233 | MGMHD, Multigrid 3-D for the Analysis of Magnetohydrodynamic (MHD) Channels |

psr-0261 | MICAP, Ionization Chamber Detector Response by Monte-Carlo |

nea-1562 | MICROX-2, Group Constant Generator with Resonance Interference and Self-Shielding |

nea-0388 | MIGROS, Group Cross-Sections, Self-Shielding Factors, Scattering Transfer, Fission from KEDAK |

nesc9460 | MILDOS-AREA, Radiological Impact of Airborne U238 from Mining and Milling |

uscd1097 | MINEQL, Chemical Equilibrium Composition of Aqueous Systems |

ests0143 | MINET, Transient Fluid Flow and Heat Transfer Power Plant Network Analysis |

nea-0639 | MINIGAL, Average Thermal Cross-Sections, Epithermal Cross-Sections, Fission Cross-Sections from UKNDL |

nesc0888 | MINPACK-1, Subroutine Library for Nonlinear Equation System |

nesc1101 | MINTEQ, Geochemical Equilibria in Ground Water |

psr-0105 | MINX, MultiGroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX |

nea-0474 | MISSIONARY, ENDF/B to UKNDL Format Conversion |

iaea1313 | MIXER2010, Cross Sections Calculations for a Composite Mixture of ENDF Format Material |

nesc0632 | MMM-3, Semi Rigid Molecule Normal Modes and Frequencies for Slow Neutron Scattering Calculation |

nea-1706 | MMRW, Canadian and early British Energy Reports on Nuclear Reactor Theory (1940-1946) |

nea-1792 | MMRW-BOOKS, Legacy books on slowing down, thermalization, particle transport theory, random processes in reactors |

ccc-0841 | MMS3D, Method of Manufactured Solutions for 3D one-group SN Equations with escalating order of non-smoothness |

nea-1005 | MOBIDIC, Fast Reactor Hexagonal Infinite Lattice 2 Component Fuel Pin Diffusion Coefficient |

iaea1238 | MOCA, Criticality of VVER Reactor Hexagonal Fuel Assemblies |

psr-0365 | MOCUP, MCNP/ORIGEN Coupling Utility Programs |

nesc0653 | MOCUS, Minimal Cut Sets and Minimal Path Sets from Fault Tree Analysis |

nesc0491 | MOD-5, Time-Dependent MultiGroup Slowing-Down Neutron Spectra and Keff Calculation, Green Function Method |

nea-0540 | MODESTY, Statistical Reaction Cross-Sections and Particle Spectra in Decay Chain |

nea-1279 | MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors |

nea-1762 | MODLIB, library of Fortran modules for nuclear reaction codes |

nea-1414 | MOLGEO, Molecular Structure Data Tables |

nea-0527 | MONK, Keff, Collision Rate, Flux Distribution in General Geometry from UKNDL by Monte-Carlo Method |

nea-1747 | MONTE-CARLO-WS-2005, Proceedings of Monte Carlo Criticality Calculations & TRIPOLI-IV Workshops 2005 |

psr-0455 | MONTEBURNS 2.0: An Automated, Multi-Step Monte Carlo Burnup Code System |

psr-0411 | MORECA, Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup |

ccc-0127 | MORSE, MultiGroup Neutron Transport and Gamma Transport for Complex Geometry Shields by Monte-Carlo |

ccc-0431 | MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method |

ccc-0474 | MORSE-CGA, Monte-Carlo Neutron and Gamma MultiGroup Transport with Array Geometry |

nea-1181 | MORSE-DD, Monte-Carlo Neutron Transport with Combinatorial Geometry Using DDL Cross-Sections Library |

ccc-0127 | MORSE-E, Program MORSE with Uniform Source for Various Geometry |

ccc-0588 | MORSE-EMP, Monte-Carlo Neutron and Gamma MultiGroup Transport with Array Geometry, for PC |

psr-0142 | MORSEC-SP, Step Function Angular Distribution for Cross-Sections Calculation by Program MORSE |

nesc0678 | MORTRAN-2, FORTRAN Language Extension with User-Supplied Macros |

nea-1633 | MOSRA-LIGHT, High Speed 3-D X-Y-Z Nodal Diffusion Code for Vector Computers |

nea-1896 | MOSRA-SRAC, Lattice Calculation Module of the Modular Code System for Nuclear Reactor Analyses MOSRA |

nesc0551 | MOXY-MOD32, Thermal Analysis Swelling and Rupture of BWR Fuel Elements During LOCA |

nesc0551 | MOXY/MOD-1, Thermal Analysis Swelling and Rupture of BWR Fuel Elements During LOCA |

ests1098 | MPICH, Message Passing Interface (MPI) Subroutine Library for Parallel Computers and Networks |

nesc0798 | MSF21/VTE21, Desalination Plant Heat, Mass Balance, Design, Cost Optimization |

iaea1349 | MSM-SOURCE, Neutron Source Generator for MCNP from Proton Neutron Interaction |

nesc0508 | MUCHA1, Fuel Rod Pair Thermohydraulics During LOCA and ECCSA for LWR |

nesc0508 | MUCHA2, Primary Coolant Thermohydraulics During LOCA and ECCS for LWR |

nea-0933 | MULTI-KENO, Criticality Safety Analysis by Monte-Carlo |

nea-1041 | MULTIPLET, Large Event Trees for Risk Assessment Calculation |

iaea0907 | MUP-2, Fast Neutron Nuclear Reaction Cross-Sections of Medium-Heavy Nuclei |

nea-0035 | MUPO, Critical 43 Group Spectra Calculation for Homogeneous Reactor |

nea-1845 | MURE v2 - SMURE, MCNP Utility for Reactor Evolution: couples Monte-Carlo transport with fuel burnup calculations |

iaea0890 | MURLI, 1-D Flux, Reaction Rate in Cylindrical Geometry Thermal Reactor Lattice by Transport |

iaea0892 | MURLI-CLUSTER, Lattice Calculation of PHWR with Rod Clustered Fuel |

nea-1451 | MUTIL, Asymmetry Factor of Mott Cross-Sections for Electron, Positron Scattering |

nea-1673 | MVP/GMVP V.3, MC Codes for Neutron & Photon Transport Calc. based on Continuous Energy and Multigroup Methods |

iaea1411 | NAAPRO, Neutron Activation Analysis Prognosis and Optimization code |

ccc-0164 | NAC, Neutron Activation Analysis and Isotope Inventory |

nea-0806 | NAIAD, LOCA Transient and Steady-State 2 Phase Flow in Channel Network |

psr-0085 | NAISAP, Theory and Use of Gamma-Ray Spectrum Analysis Codes for NaI(Tl) Detectors. |

iaea0863 | NANICK, Infinitely Dilute Group Constant and Scattering Matrix from ENDF/B |

nesc9644 | NASA-VOF2D, 2-D Transient Free Surface Incompressible Fluid Dynamic |

nesc9568 | NASA-VOF3D, 3-D Transient, Free Surface, Incompressible Fluid Dynamic |

nesc0719 | NATRAN-2, LMFBR Piping System Pressure Transients, Fluid Hammer and Na H2O Reaction |

nesc0718 | NATRANSIENT, LMFBR Piping System Pressure Transients, Fluid Hammer, Na H2O Reaction |

nea-0853 | NAUA-MOD5 NAUA-MOD5/M, Aerosols in Reactor Containment During Meltdown |

ccc-0462 | NCRP49, X-Ray Shielding for Radiographic and Fluoroscopic Diagnostic Units |

nea-0599 | NE-SPEC, Ne-213 Liquid Scintillation Detector Fast Neutron Spectra Unfolding |

nesc0171 | NEARREX, Compound Nucleus Neutron Cross-Sections |

nea-1158 | NEARSOL, Aqueous Speciation and Solubility of Actinides for Waste Disposal |

iaea1173 | NEHEX-3D, 3-D Neutron Diffusion for Fast Reactors and WWER in Hexagonal Geometry |

nea-1422 | NESKA, Electron and Positron Scattering from Point Nuclei |

ccc-0641 | NESTLE, Few-Group Neutron Diffusion for Steady-State and Transient Problems by Nodal Expansion Method (NEM) |

nea-0823 | NEUPAC, Experimental Neutron Spectra Unfolding with Sensitivities |

nea-1635 | NIRAD, A Two-Dimensional Radiation Hydrodynamics Code |

ccc-0582 | NITRAN, Neutron Transport Code System Based on Anisotropic Scattering |

nesc0709 | NIXLIN, Function Minimization Using Direct Search Simplex Method for Nonlinear Equation Fit |

nea-1025 | NJOY-UTILITIES-EIR, Utility Program EPLOTR, CPLOTR, SEPR, COMBR, DECAYR for NJOY |

psr-0480 | NJOY99, Data Processing System of Evaluated Nuclear Data Files ENDF Format |

ests1365 | NLCGCS_MPV3.0, Inversion of electromagnetic fields for subsurface electrical properties |

nea-0974 | NMTC/JAERI97, High-Energy P, N, Pion Reaction Monte-Carlo Simulation |

nea-1653 | NMTC/JAM, Simulates High Energy Nuclear Reactions and Nuclear-Meson Transport Processes |

uscd1018 | NONSAP, Finite Element Calculation for Nonlinear Static and Dynamic Analysis of Complex Structures |

nesc0974 | NONSAP-C, Static and Dynamic Loads of 3-D Reinforced Concrete Structures |

nea-0671 | NORCOOL, BWR LOCA Analysis with Thermal Non-Equilibrium and Counter Current Flow |

ests0262 | NORIA, 2-D Non-Isothermal 2-Phase Flow Through Porous Media |

nea-1388 | NORMA, Neutron & Thermo-Hydraulic Behaviour of LWR's by Coarse-Mesh Diffusion Methods |

nea-1611 | NORMA-FP, Perform Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions |

nea-0921 | NOTAM, Neutronics Hydraulics of BWR in Steady-State Conditions |

iaea1171 | NOTRAN/3D, 3-D Neutron Transport in X-Y-Z Geometry by Discrete Nodal Transport Method |

ccc-0684 | NRCDOSE 2.3.20, Evaluation of Routine Radioactive Effluents from Nuclear Power Plants |

ccc-0768 | NRCDOSE72 1.2.3, Evaluation of Routine Radioactive Effluents from Nuclear Power Plants with Windows Interface |

ests1049 | NRCPIPES, Fracture Mechanics of Cracked Pipes |

nea-0700 | NRESP-3, Organic Scintillation Detector Response to Monoenergetic Fast Neutron |

nea-0125 | NRN, Removal-Diffusion for Squares and Cylindrical Geometry with Energy Transfer Matrix |

iaea1389 | NRSC, Neutron Resonance Spectrum Calculation System |

nea-1347 | NSLINK, Coupling of NJOY Cross-Sections Generator Code to SCALE-3 System |

nesc0790 | NUBOW-2D/INEL, 2-D Core Restraint System Stress Analysis, with Bowing, Creep, Swelling |

nea-0951 | NUCCON, Nuclide Concentration and Activation in D-T Fusion Reactor |

iaea1320 | NUCHART, Nuclear Properties and Decay Data Chart |

nea-1492 | NUCLEUS-CHART, Interactive Chart of Nuclides |

nesc0683 | NUFUEL, Conditions for Power Production, U Fuel, Pu Recycle and Reprocessing |

nesc9888 | NUTRAN, Doses by Radionuclide Migration from Nuclear Waste Storage |

iaea0918 | NX-1, Excitation Function of (N-P) and (N-He4) Reaction |

iaea0919 | NX-2, Excitation Function of (N-D) and (N-He3) Reaction |

uscd1242 | NucWiz, set up and run Monte Carlo calculations |

psr-0014 | O5S, Calibration of Organic Scintillation Detector by Monte-Carlo |

nesc1125 | OCA-P, PWR Vessel Probabilistic Fracture Mechanics |

nesc0898 | OCTAVIA, PWR Pressure Vessel Failure Probability for Routine Pressure Transients |

uscd1232 | ODEPACK, Initial Value Problems of Ordinary Differential Equation System |

ccc-0046 | OGRE, Monte-Carlo System for Gamma Transport Problems |

nea-1591 | OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo |

nea-1271 | OMICRON, LLNL ENDL Charged Particle Data Library Processing |

ccc-0266 | ONETRAN, 1-D Transport in Planar, Cylindrical, Spherical Geom. for Homogeneous, Inhomogeneous Probl., Anisotropic Source |

nea-1877 | OPBA, Operator Procedural Behavior Analyzer |

nea-0552 | OPTIM, Minimization of Band-Width of Finite Elements Problems |

nesc0829 | OPTIMIZERS, Subroutine Library for Unconstrained Nonlinear Optimization Problems |

iaea1316 | OPTMOD, Elastic and Total Cross-Sections, Polarization by Optical Model |

nesc0703 | ORCENT-2, Full Load Steam Turbine Cycle Thermodynamics for LWR Power Plant |

nesc0588 | ORCOST-2, PWR, BWR, HTGR, Fossil Fuel Power Plant Cost and Economics |

nea-1324 | OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN |

ccc-0371 | ORIGEN-2.2, Isotope Generation and Depletion Code Matrix Exponential Method |

ccc-0702 | ORIGEN-ARP 2.00, Isotope Generation and Depletion Code System-Matrix Exponential Method with GUI and Graphics Capability |

nea-0622 | ORIGEN-JR, Radiation Source and Nuclide Transmutation with In-Core Burnup |

nea-1249 | ORION-II, Concentration and Dose from Radioactive Release into Atmosphere |

nea-1880 | ORIP-XXI, isotope transmutation simulations |

ests0329 | ORMGEN3D, 3-D Crack Geometry FEM Mesh Generator |

psr-0275 | ORMONTE, Uncertainty Analysis for User-Developed System Models |

nesc0525 | ORTHAT, Transient Heat Conduction in 2-D X-Y, R-Z and R-Theta Geometry |

nesc0525 | ORTHIS, Steady-State Heat Conduction in 2-D X-Y, R-Z and R-Theta Geometry |

nesc1102 | ORTURB, HTGR Steam Turbine Dynamic for FSV Reactor |

nesc9469 | OTTER, Resolution Style Theorem Prover |

nea-0802 | OWEN-1, LOCA Transient and Steady-State 2 Phase Flow in Heated Channel |

psr-0538 | P-CARES 2.0.0, Probabilistic Computer Analysis for Rapid Evaluation of Structures |

nesc0926 | PABLM, Doses from Radioactive Releases to Atmosphere and Food Chain |

nesc0901 | PAD, Coupled Neutronics, Thermohydraulics in 1-D Spherical, Cylindrical, Planar Geometry |

ccc-0621 | PAGAN-1.1, Low-Level Nuclear Waste in Ground Water, Performance Assessment Code |

nea-1008 | PALLAS-1D(VII), Direct Integration of Transport Equation in 1-D Planar and Spherical Geometry |

nea-0702 | PALLAS-2DY, 2-D Neutron Transport, Gamma Transport with Anisotropic Scattering for Fixed Source |

psr-0156 | PAPIN, Cross Section, Self-Shielding Factors for Fertile Isotopes in Unresolved Resonance Region |

nesc0555 | PARET-ANL(NESC), Thermohydraulics of Reactivity Accident in LWR |

psr-0516 | PARET-ANL, Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores |

ccc-0499 | PART61, Low Level Radioactive Waste Impact Analysis |

ccc-0760 | PARTISN 5.97, 1-D, 2-D, 3-D Time-Dependent, Multigroup Deterministic Parallel Neutral Particle Transport Code |

nea-0521 | PAS-1, 2-D, 3-D Linear Static and Dynamic Stress Analysis with 2-D Steady-State Temperature Distribution |

nea-1238 | PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation |

nea-1680 | PASCAL, Probabilistic Fracture Mechanics Analysis of Structural Components in Aging LWR |

ccc-0445 | PAVAN, Atmospheric Dispersion of Radioactive Releases from Nuclear Power Plants |

nesc9617 | PC-BLAS, PC Linear Algebra Subroutines |

ests0071 | PC-PRAISE, BWR Piping Reliability Analysis |

nesc1057 | PCC/SRC, PCC and SRC Calculation from Multivariate Input for Sensitivity Analysis |

ests0764 | PCDOSE-ESTSC, Radioactive Dose Assessment and NRC Verification |

nesc9917 | PCHIP, Piecewise Cubic Hermite Data Interpolation |

uscd1205 | PCNUDAT-PCNULIB, Nuclear Properties Data Base and Retrieval System |

iaea1220 | PCROSS, Pre-Equilibrium Emission Spectra in Neutron Reactions |

ests1145 | PCX, Interior-Point Linear Programming Solver |

ests0847 | PDASAC, Partial Differential Sensitivity Analysis of Stiff System |

nesc9839 | PDES, Fips Standard Data Encryption Algorithm |

iaea1261 | PEGAS, Unified Model for Particle and Gamma Emission Nuclear Reactions |

nesc0865 | PELE-IC, 2-D Eulerian Incompressible Hydrodynamic and Bubble Dynamic after LWR LOCA |

iaea0819 | PELINOMIC, Power Plant Cost Optimization for Dispersed Load Centres |

iaea0829 | PELINSCA, Elastic Scattering and Total Cross-Sections and Polarization by Hauser-Feshbach |

iaea0855 | PELSHIE, Dose Rates from Gamma Source by Point-Kernel Integration |

nea-1525 | PENELOPE2014, A Code System for Monte-Carlo Simulation of Electron and Photon Transport |

nea-1886 | PENGEOM, tools for handling complex quadric geometries in Monte Carlo simulations of radiation transport |

nea-1339 | PEPIN, Methodology for Computing Concentrations, Activities, Gamma-Ray Spectra, and Residual Heat from Fission Products. |

iaea1185 | PEQAG-2, Pre-Equilibrium Model Nucleon, Gamma Spectra and Cross-Sections |

nesc9800 | PFPL, Puff Plume Atmospheric Radioactive or Toxic Deposition |

iaea1413 | PGAA-IAEA, Database for Prompt Gamma-ray Neutron Activation Analysis |

uscd1222 | PHAST, Calculation of isotope equilibrium constants for geochemical models |

psr-0432 | PHAZE, Parametric Hazard Function Estimation |

iaea1327 | PHENOM, Nuclear Level Densities of Excited Nuclei |

nea-1857 | PHITS-2.88, Particle and Heavy Ion Transport code System |

uscd1207 | PHREEQC, Modeling of Geochemical Reactions, Calculation of pH, REDOX Potential |

uscd1207 | PHREEQCI, Windows Interactive Version of PHREEQC |

nesc9674 | PHREEQE, Modelling of Geochemical Reaction, Calculation of P-H, Redox Potential |

uscd1207 | PHRQCGRF, code to create graphs from the data generated by PHREEQC |

uscd1183 | PHRQPITZ, Geochemical Calculation in Brines |

ccc-0160 | PICA, Photon-Induced Medium-Range Nuclear Cascade Analysis by Monte-Carlo |

psr-0568 | PICES, Probabilistic Investigation of Capacity and Energy Shortages |

ests0585 | PICL, Portable Instrumented Communication Library |

psr-0238 | PICTURE, 2-D Slices Through 3-D CG of MORSE, QAD-CG |

nea-1084 | PIEDEC, Effective Dose Equivalent from Inhalation or Ingestion |

nea-1612 | PIN99W, Modelling of VVER and PWR Fuel Rod Thermomechanical Behaviour |

nea-0416 | PIPE, 1-D Gamma Transport for Slab, Spherical Shields with Compton Scattering Calculation |

ests0650 | PIPE-ESTSC, Friction Factor for 3-D Turbulent Flow in Rough Tubes |

nea-0482 | PIXSE, Scattering Moments Calculation from Scattering Law |

iaea1172 | PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method |

ccc-0381 | PLACID, Gamma Streaming in Cylindrical Duct Shields by Monte-Carlo |

psr-0106 | PLASMX, MultiGroup Neutral Particle Transport in Tokamak CTR Plasma |

nesc0586 | PLENUM, Bulk Flow Distribution in Cylindrical Reactor Coolant Inlet Plenum, Potential Flow |

nesc0591 | PLETHS, Isopleth Area for Pollution Downwind from Single Steady-State Source |

nesc0544 | PLOT-3D, Graphics Subroutines for 3-D Surface Plots with Arbitrary Rotations |

nea-1879 | PLOT-S, Plotting Program with special Features for Windows Environment |

iaea0936 | PLOTC4, Plotting of ENDF/B and EXFOR Data |

uscd1215 | PLOTEF, ENDF/B Data Plot |

nea-0522 | PLOTENDF, Log-Log Plot of ENDF/B Point Cross-Sections |

nesc1130 | PLOTNFIT.4TH, Data Plotting and Curve Fitting by Polynomials |

iaea1329 | PLOTTAB, Curve and Point Plotting with Error Bars |

nea-0493 | PLUDOS, Ground Level Gamma Dose from Radioactive Release at Various Heights |

nea-0704 | PLUMEX, Gamma Doses from Atmospheric Plume |

nea-1663 | PLUTON, Isotope Generation and Depletion in Highly Irradiated LWR Fuel Rods |

nea-0464 | PN, 1-D, 2-D, 3-D MultiGroup Neutron Transport |

ests0428 | POISSON SUPERFISH, Poisson Equation Solver for Radio Frequency Cavity |

nea-1058 | POISSON, Analysis Solution of Poisson Problems in Probabilistic Risk Assessment |

nea-0488 | POISSX, Poisson Equation on Rectangle with Various Boundary Condition |

nesc0639 | POLLA-NESC, Resonance Parameter R-Matrix to S-Matrix Conversion by Reich-Moore Method |

iaea0944 | POLLA/IECTA, ENDF/B Reich-Moore to Adler-Adler Resonance Parameter Conversion |

iaea1249 | POTAUS, Stopping Power and Particle Ranges in Various Material |

nea-1675 | PPICA, Power Plant Investment Cost Analysis |

nesc1070 | PRAISE-C, Double-Ended Guillotine Break (DEGB) Breaks from Weld Cracks in Light-Water Reactor Piping System |

nesc9983 | PRAXIS, High Level Computer Language for System Applications |

nea-0809 | PREANG, Spectra and Angular Distribution from Nuclear Reaction by Statistical Model |

nea-0904 | PRECIP-2, Zircaloy Cladding Oxidation Simulation for LWR under LOCA Conditions |

psr-0226 | PRECO-2000, Exciton Model Preequilibrium Code System with Direct Reactions |

psr-0226 | PRECO-D2, Pre-Equilibrium and Direct Reaction Double Differential Cross-Sections |

nea-0509 | PREDEX-1, U, Pu, Nitric Acid Distribution in Counter Current Solvent Extraction |

nea-0888 | PREM, Pre-Equilibrium Energy Spectra and Cross-Sections for Multiple Nucleon Emission |

nesc0528 | PREP KITT, System Reliability by Fault Tree Analysis |

nea-1173 | PREP, Input Preparation for Monte-Carlo Program SPOP |

nesc0528 | PREP, Min Path Set and Min Cut Set for Fault Tree Analysis, Monte-Carlo Method |

nea-1485 | PREP-45, Input Preparation for CITATION-2 |

iaea1379 | PREPRO2010, Data Preparation and Management, Subsidiary Calculations (ENDF Format) |

nea-0251 | PREST, Pressure Temperature Transients, I Inhalation in Containment Building from LOCA |

iaea0905 | PRESTO, Slab Shields for Time-Dependent Gamma Spectra |

ccc-0504 | PRESTO-II, Low Level Radioactive Waste Transport and Risk Assessment |

nea-0695 | PROCIV, Protection Coefficient from Fallout in Residential Area Housing |

nea-0169 | PROCOPE, Collision Probability in Pin Clusters and Infinite Rod Lattices |

nea-1170 | PROF-DD, Generator of MultiGroup Cross-Sections Library DDX for MORSE-DD, ANISN-DD, DOT-DD |

nesc1023 | PROGRAM-H, Analysis of Transonic Airfoils with Turbulent Boundary Layer Correlation |

ests0790 | PROGRAM-K, Transonic Airfoil, Turbine, Compressor Blade Design |

nesc0846 | PROMSYS, Plant Equipment Maintenance and Inspection Scheduling |

iaea1216 | PRORIA, Fast Reactivity Transients in PWR with Two-Phase Flow Model |

nesc0778 | PROSA-1 PROSA-2, Accidents Probability Analysis Using Response Surface Method |

nesc0542 | PSA-2, Stress Analysis, Thermal Expansion and Loads in Multi Anchor Piping System |

nea-1138 | PSACON, Conversion Program for PSAOUT-I Output Files |

iaea1174 | PSAPACK, Probabilistic Safety Analysis with Fault Event Trees |

iaea0888 | PSU-LEOPARD, Program LEOPARD in PFMP System, Fast Neutron and Thermal Neutron Spectra Calculation |

uscd1216 | PSYCHE, ENDF/B Data Consistency Check in ENDF Format |

nesc0155 | PTH-1, Pressure and Temperature in Containment after Blowdown of H2O Coolant System |

ccc-0618 | PTRAN, Proton Transport for 50 to 250 MeV by Monte-Carlo |

psr-0157 | PUFF-2, MultiGroup Covariance Matrices from ENDF/B-5 Error Files |

psr-0534 | PUFF-IV 6.1.0, Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files |

iaea1228 | PULSTRI, Mixed Core Triga Reactor Pulse Calculation |

ccc-0595 | PUTZ, Point-Kernel 3-D Gamma Shielding |

nea-1679 | PVIS-4, Pressure vessel irradiation, source preparation |

nesc0441 | PWCOST, Fuel Cycle Cost and Economics by Present Worth Levelized Method |

nesc1081 | PWR-GALE, Radioactive Gaseous Release and Liquid Release from PWR |

nesc0552 | PWR-PPM, Boration-Dilution Tables Generator for PWR Operation |

nea-1828 | Proceedings of PHYSOR'90 conference: Physics of Reactors, Operation, Design and Computation, Marseille, 23-27 April 1990 |

ccc-0493 | QAD-CGGP, Fast Neutron and Gamma Penetration in Shields with Combinatorial Geometry |

ccc-0645 | QAD-CGGP-A, Fast Neutron, Gamma Penetration in Shields with Combinatorial Geometry |

ccc-0396 | QADMOD-G, Point-Kernel Gamma-Ray Shielding Program |

ccc-0617 | QBF, Radiation Dose Distribution Around Spent Fuel Shipping Casks |

uscd1200 | QCALC, Reaction and Decay Q-Values, Threshold Energies from Atomic Masses |

nesc0612 | QMESH RENUM QPLOT, Mesh Generator on 2-D Bodies for Finite Element Method Analysis, with Plot Utility |

ests0332 | QMESH RENUM QPLOT, Self-Organizing Mesh Generator |

nea-0819 | QUADPACK, Numerical Integration by Gauss Kronrod Quadrature |

nea-1600 | QUARK, 2-Group 3-D Neutronic Kinetics Coupled to Core Thermalhydraulics |

ccc-0556 | QUINCE, Dose Absorption, Health Risk from Skin Contamination |

nesc0474 | QX-1, 1-D MultiGroup Time-Dependent Neutron Diffusion in Planar Cylindrical and Spherical Geometry for Fast Reactors |

nesc0255 | R-101, 1 Group Space-Independent Reactor Kinetics for Neutron Density |

nesc0168 | R-102, 1 Group Space-Independent Inverse Reactor Kinetics |

nesc0281 | RABBLE, Cross-Sections from Single-Level Resonance Parameter, Homogeneous or Heterogeneous Infinite System |

ests0062 | RABFIN PARTS, Noble Gas, Iodine, Particulate Gaseous Effluent Dose Parameters |

ccc-0639 | RACC-PULSE, Neutron Activation in Fusion Reactor System |

ccc-0627 | RADAC, Radioactive Decay and Accumulation of Long Lived Isotopes |

nea-0487 | RADAK, Multichannel Analyser Neutron Spectra and Gamma Spectra Unfolding |

psr-0348 | RADCOMPT, Sample Analysis for Alpha and Beta Dual Channel Detectors |

nea-0467 | RADHEAT, Transport, Heat Generator, Radiation Damage Cross-Sections in Reactor and Shield |

nea-0181 | RADIFLUX, Bessel Function Fit of Radial Flux Data in Cylindrical Reactor |

ccc-0422 | RADRISK, Doses to Human Organs and Health Effects from Inhalation and Ingestion |

ccc-0800 | RADTRAD 3.03, Model for Radionuclide Transport and Removal and Dose Estimation |

iaea1350 | RAF, Direct Reaction Radiation Capture Cross-Sections in Giant Resonance Region |

nesc0631 | RAFFLE-2/MOD2, 3-D Steady-State Monte-Carlo Neutron Transport, Cell Averaged Scattering Cross-Sections Calculation |

ccc-0083 | RAID, Gamma, Neutron Scattering into Cylindrical or Multibend Duct |

iaea0822 | RAM-1, Thermal Flux Derivatives at Plane Geometry Control Rod Boundary by Monte-Carlo |

nea-0843 | RANCH, Radionuclide Migration in Geological Media |

nesc0843 | RANDOM_NUMBERS, Random Number Sequence Generated from Gas Ionisation Chamber Data |

nea-1867 | RAPID, RAdial power and burnup Prediction by following fissile Isotope Distribution in the pellet |

nea-1539 | RAPRAN, Radionuclide Migration from Waste Glass Release |

nea-0632 | RAPVOID, H2O Flow and Steam Flow in Pipe System with Phase Equilibrium |

ccc-0783 | RASCAL 4.2, Radiological Doses from Accidental Release to Atmosphere |

nesc0758 | RASE4, Ion Implantation in Solids, Range, Straggling, Energy Deposition, Recoils |

nea-0475 | RASPA, Burnup with Fission Products Inventory, Gamma Spectra, Isotopic Power Density |

ests0050 | RATAF, Radioactive Liquid Tank Failure |

ccc-0632 | RBD, Doses from Radionuclide Inhalation, Ingestion, Wound Uptake from Bioassays |

nesc1090 | RCSLK9, PWR Coolant System Leak Rate |

nea-0168 | RDMM, Flux Spectra from In-Pile Fast Neutron Activation Experiment |

ccc-0443 | REAC*3, Isotope Activation and Transmutation in Fusion Reactors |

nea-1873 | REACTORPHYSICS-62-91, Archive of Reactor Physics Reports and Summaries of [N]EACRP (1962-1991) |

nea-1814 | REACTORSHIELDING-NMS, Reactor Shielding for Nuclear Engineers by N. M. Schaeffer |

iaea0846 | REBEL-3, Whole Body and Organ Gamma Doses of Inhomogeneous Phantom by Monte-Carlo |

ccc-0708 | REBUS-PC 1.4, Code System for Analysis of Research Reactor Fuel Cycles |

ccc-0653 | REBUS3/VARIANT8.0, Code System for Analysis of Fast Reactor Fuel Cycles |

ests0176 | RECAP, Replacement Energy Cost for Short-Term Reactor Plant Shut-Down |

iaea0848 | RECENT2010, Reconstruction of Cross Sections Data from Resonance Parameters |

nesc9967 | RECOG-ORNL, Pattern Recognition Data Analysis |

nea-0519 | REDIFFUSION, 1-D Neutron Removal-Diffusion and Gamma Point-Kernel Calculation for Shielding |

nea-0510 | REEX-1, U, Pu, Nitric Acid Distribution in Counter Current Pluristage Stripping |

nesc1065 | REFCO83, Nuclear Fuel Cycle Cost Economics Using Discounted Cash Flow Analysis |

nea-0914 | REFIT, Multilevel Resonance Parameter Least Square Fit of Neutron Transmission, Capture, Fission & Self Indication Data |

nea-0262 | REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR |

nea-1231 | REFREP, Near-Field Model for Spent Fuel Repository |

iaea1314 | RELABEL2010, Labels FORTRAN Statements in ENDF Format Processing Programs |

nesc0369 | RELAP-4, Transient 2 Phase Flow Thermohydraulics, LWR LOCA and Reflood |

nesc0917 | RELAP-5, Transient 2 Phase Flow Thermohydraulics, LWR LOCA Accidents |

nea-0437 | RELAP-UK, Thermohydraulic Transients and Steady-State of LWR |

nea-0821 | RELAP/REFLA, Core Reflooding During PWR LOCA |

nea-0615 | RELOSS, Reliability of Safety System by Fault Tree Analysis |

ests0579 | REMIT, Radiation Exposure Monitoring and Information Transmittal System |

psr-0482 | REMIT5.1, Radiation exposure monitoring and information transmittal system |

nea-0429 | REMO, Failure Analysis of System with Reparable and Standby Components by Monte-Carlo |

nea-0101 | REP-3, Time-Dependent Xe and Sm Poisoning from Space-Dependent Flux Distribution |

ccc-0586 | REPRISK-PC, Radioactive Waste Storage Risk Assessment |

nea-0932 | RESENDD, Resonance Cross-Sections Calculation from ENDF/B-4 and ENDF/B-5 |

ccc-0786 | RESRAD 6.5, Residual Radioactive Material Guideline Implementation |

ests1225 | RESRAD-BUILD2.36, Residual Radioactive Material Guideline Implementation |

iaea1286 | RETRAC, Reactor Core Accident Simulation |

nea-0979 | RETRANS, Reactivity Transients in LWR |

iaea0935 | REX1-87, MultiGroup Neutron Cross-Sections from ENDF/B |

iaea0965 | RGENDF, Conversion of NJOY MultiGroup Cross-Sections to ENDFB-5, EXPANDA, PFCOND, COMPAR Format |

nea-0508 | RHFPPP, SCF-LCAO-MO Calculation for Closed Shell and Open Shell Organic Molecules |

ccc-0137 | RIBD, Fission Products Inventory and Delay Heat in Fast Reactors, with Data Library |

ccc-0382 | RIBD-IRT, Isotope Buildup and Isotope Decay from Fission Source |

nea-0239 | RIBOT-5, 0-D Burnup for 5 Group BWR or PWR Lattice |

nesc0453 | RICE, Energy Exchange Matrix, Damage Cross-Sections, Recoil Energy Spectra from ENDF/B |

nea-0589 | RICE-CEGB, Long-Term Actinides and Fission Products Inventory of Irradiated Fuel |

nesc0720 | RICE-LASL, Hydrodynamic of Chemically Reactive Mixture by 2-D Navier Stokes Equation |

nesc9580 | RICKI, Interactive Gamma Spectra Unfolding with Isotope Identification |

nea-0234 | RICM, Resonance Absorption in Multi-Region Slab or Square or Hexagonal Lattice |

nesc0213 | RIFF-RAFF, Resonance Integrals in 2 Region Cell, Isotropic Flux and Isotropic Scattering |

nesc0638 | RIGEL-4, ENDF/B Utility, Data Retrieval, BCD to BIN Conversion |

nea-1825 | RIMACS, Reactor Inspection Main Control System |

nea-1356 | RIPP2, H2O Chemistry File Generator for Program PHREEQE |

ests0185 | RIPPLE, Incompressible Fluid Dynamics with Free Surfaces |

ccc-0486 | RISKAP, Risk Assessment of Radiation Exposure for Population |

ccc-0623 | RISKIND, Radiological Risk Assessment for Spent Nuclear Fuel Transportation |

ccc-0626 | RIVER-RAD, Radionuclide Transport in Surface Waters |

nesc0831 | RO-75, Reverse Osmosis Plant Design Optimization and Cost Optimization |

nea-1449 | ROLAIDS-CPM, 1-D Slowing-Down by Collision Problems Method |

nesc0265 | RSAC, Gamma Doses, Inhalation and Ingestion Doses, Fission Products Inventory after Fission Products Release |

ests0608 | RSAC-6, Gamma doses, inhalation and ingestion doses, fission products inventory after fission products release |

ccc-0761 | RSAC-7.2, Gamma doses, inhalation and ingestion doses, fission products inventory after fission products release |

nea-0598 | RSYST, Modular System for Reactor Core and Shielding Problems |

nesc0245 | RTS, Non-Equilibrium Reactor Kinetics in Delayed Neutron Regime |

nea-1835 | Reactor Shielding Design Manual by Rockwell T. III |

nea-0484 | S1CALC, Scattering Law for Delta Function or Gaussian Phonon Spectra |

nea-0402 | SABINE-3, Neutron Penetration and Gamma Penetration in Reactor Shield for Planar, Spherical, Cylindrical Geometry |

psr-0242 | SABRINA, Geometry Plot Program for MCNP |

nea-1884 | SACALC-ELLIPSOID, Calculates the average solid angle subtended by a ellipsoid solid or surface |

nea-1688 | SACALC3, Calculates the average solid angle subtended by a volume |

nea-1078 | SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System |

ccc-0517 | SADDE-MOD1, Beta Spectra Evaluation Input Generator for Program VARSKIN |

psr-0573 | SAEROSA, Single-Species Aerosol Coagulation and Deposition with Arbitrary Size Resolution |

nea-0460 | SAFE-2D/FBM, Elastic Stress Analysis of Mix of Plane and Axial Structure |

nesc0332 | SAFE-3D, Stress Analysis of 3-D Composite Structure by Finite Elements Method |

nesc0251 | SAFE-AXISYM, Stress Analysis of Axisymmetric Composite Structure by Finite Elements Method |

nesc0451 | SAFE-CRACK, Viscoelastic Analysis of Plane and Axisymmetric Concrete System, Finite Elements Method |

nesc0300 | SAFE-CREEP, Viscoelastic Analysis of Concrete Structure, Age Temperature and Temperature Dependent Creep |

nesc0250 | SAFE-PCRS, Stress Analysis of Axisymmetric Composite Structure by Finite Elements Method |

nesc0252 | SAFE-PLANE, Stress Analysis of Planar Structure by Finite Elements Method |

nesc0253 | SAFE-SHELL, Stress Analysis of Axisymmetric Thin Shells by Finite Elements Method |

nesc0674 | SAFTAC, Monte-Carlo Fault Tree Simulation for System Design Performance and Optimization |

nea-1779 | SAGEP-FR, Sensitivity Analysis of Fast Reactor Parameters |

nea-0212 | SAHYB-2, Solution of Ordinary Differential Equation with User-Supplied Subroutine |

nesc0919 | SALE, Quality Control of Analytical Chemical Measurements |

nesc1069 | SALE-3D, 3-D Fluid Flow, Navier Stokes Equation Using Lagrangian or Eulerian Method |

iaea0861 | SALLY, Dynamic Behaviour of Reactor Cooling Channel by Point Model |

nesc9849 | SALT-4, Temperature and Stress from Radioactive Waste Disposal in Bedded Salts |

ccc-0187 | SAM-CE, Time-Dependent 3-D Neutron Transport, Gamma Transport in Complex Geometry by Monte-Carlo |

nesc1120 | SAMCR, 2-D Elastodynamic Fracture Analysis |

psr-0158 | SAMMY 8.1.0, Multilevel R-Matrix Fits to Neutron and Charged-Particle Cross-Section Data Using Bayes' Equations |

nesc0879 | SAMPLE, Mean and Standard Deviation and Probability of Given Function by Monte-Carlo |

nea-0691 | SAMPO80, Ge(Li) Detector Gamma Spectra Unfolding with Isotope Identification |

iaea0837 | SAMSY, Neutron and Gamma Dose Rates and Heat Source for Multilayer Shields |

nesc9603 | SANCHO, Quasistatic Large Deformation Inelastic Response of Planar, Axial Solids |

ccc-0112 | SAND-2, Neutron Flux Spectra from Multiple Foil Activation Experiment |

ccc-0361 | SANDYL, 3-D Time-Dependent and Space-Dependent Gamma Electron Cascade Transport by Monte-Carlo |

nesc0641 | SAP-4, Static and Dynamic Linear System Stress Analysis for Various Structures |

psr-0608 | SAPHIRE 8.0.9, Systems Analysis Programs for Hands-On Integrated Reliability Evaluations |

nea-0520 | SARAZE-2, Energy Release from Reactivity Transient Fast Reactor Accident |

nea-0204 | SASSI, Total and Differential Elastic and Inelastic Neutron Cross-Sections by Hauser-Feshbach |

nea-1898 | SAUNA V1.1, Severe Accident UNcertainty Analysis |

iaea0917 | SC2N3N, (n-2n) and (n-3n) Cross-Sections Systematics |

ccc-0834 | SCALE 6.2.3, A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design |

nea-1405 | SCALPLO, Plotting of Flux Output from SCALE Program |

psr-0352 | SCAMPI, Problem Dependent Library Preprocessing in AMPX Format |

nesc1119 | SCANS, Shipping Cask Design Safety Analysis |

ccc-0418 | SCAP-82, Single Scattering, Albedo Scattering, Point-Kernel Analysis in Complex Geometry |

nea-0444 | SCARF-4, Nonlinear Stresses in Pressure Vessel Liner with Plastic Behaviour Simulation |

nea-0829 | SCAT-2, Cross Sections and Angular Distributions for Spherical Nuclei by Optical Model |

nea-0829 | SCAT-2B, Spherical, Optical Model Cross Sections Calculation for N, P, D, T, He3, He4, Heavy Ions |

iaea0913 | SCENARIOS, Simulation of Reactor Introduction and Operation Scenario Needs |

nea-0431 | SCEPTIC, Pressure Drop, Flow Rate, Heat Transfer, Temperature in Reactor Heat Exchanger |

ccc-0826 | SCEPTRE 1.7, Sandia Computational Engine for Particle Transport for Radiation Effects |

nesc0802 | SCHAFF, Single-Phase Flow, Heat Transfer in Porous Media, Geothermal Energy System |

psr-0267 | SCINFUL, Scintillation Neutron Detector Response by Monte-Carlo |

nea-1755 | SCIP, Radioactive Surface Contamination Investigation Program |

psr-0210 | SCOPE, Shipping Cask Optimization and Parametric Evaluation |

nea-0235 | SCOPERS-2, BWR and PWR Core Performance Simulation |

nea-0498 | SCORCH-B2, BWR Core Heating During LOCA |

nea-0407 | SCORE-4, 2-D Removal Diffusion in X-Y or R-Z Geometry for Rectangular Shields |

ests0015 | SCORE-EVET, 3-D Hydraulic Reactor Core Analysis |

nea-0537 | SCOTCH, 1-D 2 Group HTGR Core Kinetics with Temperature Transients and Fluid Dynamic |

nesc1002 | SCREEN, Statistical Sensitivity Ranking of Program Input Variables |

nea-1540 | SCRELA, LOCA Analysis of Super-Critical Light-Water Reactors |

nea-0865 | SCRIMP, Steady-State Thermohydraulics of HTGR Subchannel |

nesc9717 | SCWEB, Scientific Workstation Evaluation Benchmark |

ccc-0620 | SEECAL-2.0, Specific Effective Energy in Human Body Due to Radiation |

nesc1063 | SEISIM-1, Seismic Probabilistic Risk Assessment |

nea-0654 | SELFS-3, Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-2 |

nesc9438 | SENSIT MUSIG COMSEN, Sensitivity Test Analysis |

ccc-0405 | SENSIT, Integral Response Sensitivity from Neutron Cross-Sections, Gamma Cross-Sections Errors |

ccc-0729 | SERA-1C1, Simulation environment for radiotherapy applications |

nea-1840 | SERPENT 1.1.7, 3-D continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications |

ccc-0629 | SESOIL, 1-D Vertical Transport for Unsaturated Soil Zone |

uscd1217 | SETMDC: Preprocessor for CHECKR, FIZCON, INTER, etc. ENDF Utility source codes |

nesc0623 | SETS, Boolean Manipulation for Network Analysis and Fault Tree Analysis |

ccc-0310 | SFACTOR, Dose Equivalent to Target Organs from Radionuclides in Organs |

iaea0841 | SFAK, Unscattered Gamma Self-Absorption from Regular Fuel Rod Assemblies |

nea-1239 | SFERXS, Photoabsorption, Coherent, Incoherent Scattering Cross-Sections Function for Shielding |

iaea1356 | SGNUCDAT, Nuclear Data Display for IAEA Safeguard Material Analysis |

nea-0370 | SHADOK-3-6, Transport Equation with Anisotropic Diffusion in P1 Approximation for Spherical and Cylindrical Geometry |

nesc0893 | SHAFT-79, 2 Phase Flow in Porous Media for Geothermic Energy System |

iaea0925 | SHARDA, Thermal Reactor Isotope Irradiation Analysis |

ests0204 | SHC, Seismic Hazard Assessment for Eastern US |

nesc0452 | SHELL-5, Elastic Stress Analysis of 3-D Thin Shells Using Finite Elements Method |

iaea1287 | SHIELD, Monte-Carlo Code for Simulating Interaction of High Energy Hadrons with Complex Macroscopic Targets |

iaea1391 | SHIELDER, Gamma shielding calculations of radionuclides emitting photons 0-5 to 10 MeV |

ccc-0379 | SHIELDOSE, Doses from Electron and Proton Irradiation in Space Vehicle Al Shields |

nea-0538 | SHOSPA-MOD, Hot Spot Factors for Fuel and Clad, Hot Channel Factors |

nea-0466 | SHREDI, Neutron Flux and Neutron Activation in 2-D Shields by Removal Diffusion |

nea-0852 | SICOS, 2-D Time-Dependent Creep Calculation of Plane or Axisymmetric Concrete Structure |

iaea1416 | SIGACE, Code for Doppler broadening of ACE-formatted files |

ccc-0118 | SIGMA/B, Doses in Space Vehicle for Multiple Trajectories, Various Radiation Source |

iaea0854 | SIGMA1-2010, Doppler Broadening ENDF Format Linear-Linear. Interpolated Point Cross Section |

nea-0571 | SIGMARZ, Stress Analysis of Axisymmetric or Plane Structures |

nesc1082 | SIGPI, Probabilistic System Performance by Fault-Tree Analysis |

ests0238 | SIMION, Electrostatic Lens Analysis and Design |

nesc9593 | SIMPLE, 2-D Hydrodynamic, Heat Flow Benchmark |

nea-0319 | SIMPLED-4, 1-D Neutron Diffusion in Spherical, Cylindrical, Planar Geometry from JAERI Fast-Set and ABBN |

ests0767 | SIMSOL, Multiphase Fluid and Heat Flow in Porous Media |

psr-0139 | SIOB, Least Square Fit of Neutron Transmission Data Using Multilevel Breit-Wigner |

nesc0687 | SITE-2, Power Plant Siting, Cost, Environment, Seismic and Meteorological Effects |

nea-1570 | SITE-94, Biosphere Model for SKI Project on the island of Aspro |

nea-0770 | SITO, Environmental Impact of Major Industrial Activities |

iaea1283 | SIXPAK2010, ENDF Format Double Differential Cross Section Converter to Single Differential Format |

nea-0905 | SIXTUS-2, 2-D MultiGroup Diffusion in Hexagonal Geometry with Intranodal Solution |

nea-1426 | SIXTUS-3, 3-D Nodal Neutron Diffusion Criticality in Hexagonal Geometry |

nea-1577 | SKETCH-N 1.0, Solve Neutron Diffusion Equations of Steady-State and Kinetics Problems |

ccc-0289 | SKYSHINE, Dose Rate Outside Concrete Steel Building from 6 MeV Gamma by Monte-Carlo |

ccc-0646 | SKYSHINE-KSU, Gamma Skyshine Doses by Integral Line-Beam Method |

nesc0581 | SLADE-D, Transient Dynamic Response of Elastic Shells by Finite Elements Method |

nesc9776 | SLAP, Large Sparse Linear System Solution Package |

nea-1081 | SLAROM, Neutron Flux Distribution and Spectra in Lattice Cell |

ests0181 | SLATEC-4.1, Subroutine Library for Solution of Mathematical Problems |

nesc9770 | SLIB77, Source Library Data Compression and File Maintenance System |

ccc-0704 | SLIDERULE 1.0,Slide Rule for direct radiation exposure approximation in criticality accidents |

nesc1077 | SMACS, Probabilistic Seismic Analysis Chain with Statistics |

nea-1767 | SMAFS, Steady-state analysis Model for Advanced Fuelcycle Schemes |

nea-1046 | SMART, Radiation Dose Rates on Cask Surface |

ccc-0602 | SMART-BNL, Offsite Radionuclide Air Concentration from Reactor Accident |

nea-0026 | SMOG, Optical Model Neutron Cross-Sections with Fox-Goodwin Integral Method |

nea-0430 | SNAP, MultiGroup 3-D Neutron Diffusion in X-Z, R-Theta-Z, Hexagonal-Z, Triangular-Z Geometry |

psr-0345 | SNL-SAND-II, Neutron Flux Spectra from Multiple Foil Activation Analysis |

nesc0521 | SOCOOL-2, Molten Materials Na Coolant Interaction, Temperature and Pressure Transient |

nesc0559 | SOFIRE-2, Containment Temperature and Pressure During Na Pool Fire, 1-Cell or 2-Cell Analysis |

nesc0832 | SOLA-DF, Time-Dependent 2-D 2 Phase Flow, Eulerian Method with Various Boundary Conditions |

nesc0723 | SOLA-ICE, Compressible Fluid Flow Transients, 2-D Planar, Cylindrical Geometry, Eulerian Method |

nesc0859 | SOLA-LOOP, Transient 2 Phase Flow in Networks of 1-D Components |

nesc0651 | SOLA-SURF, 2-D Plane, Axisymmetric, Incompressible Flow Navier Stokes Equation for Transient |

nesc0948 | SOLA-VOF, 2-D Transient Hydrodynamic Using Fractional Volume of Fluid Method |

nesc9944 | SOLGASMIX-PV, Chemical System Equilibrium of Gaseous and Condensed Phase Mixtures |

nea-1826 | SOLTRAN, solving multi-dimensional simplified P2 transport and diffusion problems of hexagonal geometry in fast reactors |

uscd1100 | SOLUPLOT, Eh-pH Diagram, a02-pH Diagram Plots for Aqueous Chemical Systems |

nesc0662 | SOLVEX, Dynamic and Steady-State Mixer-Settler and Centrifugal Contactor Behaviour |

nea-1641 | SONATINA, Predicts Behaviour of Prismatic HTGR Core under Seismic Excitation |

psr-0174 | SORA, Radionuclide Analysis Data Storage and Retrieval |

nea-0187 | SOREX-1, Worst Accident Simulation in Sora Pulsed Fast Reactor |

nea-0450 | SOTHIS, PWR Fuel Cycle Equilibrium Cost Evaluation |

ccc-0661 | SOURCES-4C, Calculating Alpha, N, Fission, Delayed Neutron Sources and Spectra |

ccc-0120 | SPACETRAN, Radiation Leakage from Cylinder with ANISN Flux Calculation |

iaea0895 | SPAGAF, PWR Fuel, Cladding Behaviour with Fission Products Gas Release |

nea-0219 | SPANDE, Stress Analysis of General Spaceframe and Pipework |

ccc-0228 | SPAR, High-Energy Muon, Pion, Heavy Ion Stopping-Powers and Ranges |

ccc-0148 | SPARES, Program System for Space Radiation Environment and Shielding System Evaluation |

nea-0468 | SPARK, Time-Dependent 1-D, 2-D, 3-D Diffusion with Heat Transfer and Feedback |

nea-0219 | SPATAM, Tilt Angle Calculation of Framework for Program SPANDE |

iaea1332 | SPEC, Neutron and Charged-Particle Reactions by Optical Model, Evaporation Model |

nesc9641 | SPECFUN1, Portable Special FORTRAN Routines with Test Drivers |

psr-0263 | SPECTER-ANL, Neutron Damage for Material Irradiation |

iaea1433 | SPECTRA2010, Convert model and general tabulation to linearized spectra (MF=5) |

nea-1165 | SPEEDI,EXPRESS, Radiation Dose from Plume Release in Nuclear Accident |

nea-0374 | SPES, Fuel Cycle Optimization for LWR |

nea-0548 | SPIRIT, Plot of Geometry and Results of 2-D Finite Elements Calculation |

ests0054 | SPIRT, Stress Strains from Transient Pressure |

nea-0462 | SPLINE, Spline Interpolation Function |

nea-0609 | SPLOSH-3, 1-D Time-Dependent Coupled Neutron Kinetics Thermohydraulics for PWR Transient |

nesc9736 | SPLPKG WFCMPR WFAPPX, Wilson-Fowler Spline Generator for Computer Aided Design And Manufacturing (CAD/CAM) Systems |

nea-0157 | SPM-046, Reactor Kinetics by 1 Group Diffusion Calculation in R-Z Geometry |

nea-1173 | SPOP-4, Uncertainty and Sensitivity Analysis Monte-Carlo Program with Input from PREP |

ccc-0460 | SPOT1, Gamma-Ray Dose Rate from Cylindrical Source Volume |

nesc0279 | SPOTS, Library Generator for Program LEOPARD from Cross-Sections Data |

nesc0716 | SPRAY-3, Thermodynamics and Heat Transfer of Na Sprays in LMFBR after Pipe Failure |

psr-0266 | SPUNIT, Multisphere Neutron Spectra Unfolding |

nea-0414 | SQUID-360, 2-D Neutron Diffusion in X-Y and R-Z Geometry with Criticality Search and Constant Neutron Source |

psr-0533 | SQUIRT 1.1, predicts leakage rate and crack area for cracked pipes in nuclear power plants |

ests1052 | SQUIRT, Seepage in Reactor Tube Cracks |

nea-0842 | SRAC-95, Cell Calculation with Burnup, Fuel Management for Thermal Reactors |

nea-0919 | SRIM-2008, Stopping Power and Range of Ions in Matter |

iaea1382 | SRNA-2K5, Proton Transport Using 3-D by Monte Carlo Techniques |

nea-0684 | SSYST, Modular System for Transient Fuel Rod Behaviour Under Accident Condition |

nesc9850 | STAFAN, Fluid Flow, Mechanical Stress in Fractured Rock of Nuclear Waste Repository |

nea-1725 | STAMPI, Application to the Coupling of Atmosphere Model (MM5) and Land-surface Model (SOLVEG) |

uscd1218 | STANEF, ENDF/B Book-keeping Operations for ENDF Format Files |

iaea0971 | STAPRE-H95, Evaporation and Pre-Equilibrium Model Reaction Cross-Sections Calculations |

nea-0461 | STAPREF, Nuclear Reactions Cross-Sections by Evaporation Model, Gamma-Cascades |

iaea0882 | STAR, Fuel Management of BWR |

psr-0330 | STARCODES, Stopping Power and Ranges for Electrons, Protons, He |

nea-0986 | STATCAT, Statistical Analysis of Parametric and Non-Parametric Data |

nea-0908 | STATISTICS, Program System for Statistical Analysis of Experimental Data |

nesc9749 | STATLIB, Interactive Statistics Program Library of Tutorial System |

nea-0352 | STAX-2, Neutron Scattering Cross-Sections by Optical Model and Moldauer Theory with Hauser-Feshbach |

psr-0113 | STAY-SL, Dosimetry Unfolding with Activation, Dosimetry, Flux Error Calculation |

nea-0055 | STDY-3, Steady-State Parallel Channel Thermal Analysis of PWR |

nea-0703 | STEADY-ACE, 3-D Neutronics and Multichannel Thermohydraulics Analysis of BWR |

nesc0487 | STEAM-67, Thermodynamics Properties of H2O and Steam from ASME Tables (1967) |

nea-0575 | STESTA, Steady-State State-Variable Profiles of Thermohydraulic Piping System |

nesc9852 | STFLO, Steady-State H2O Flow in Porous Media |

nea-0549 | STIGMA STIG STEGT STAGT STABA, Stress Analysis of Dragon HTR Graphite Structure |

iaea0900 | STOFFEL-1, Steady-State In-Pile Behaviour of Cylindrical H2O Cooled Oxide Fuel Rod |

iaea0970 | STOPOW, Stopping Power of Fast Ions in Matter |

ccc-0067 | STORM, Radiation Hazard of Solar Flares for Space Vehicles |

nea-0993 | STRADE, Stratified Random Design for Reactor Safety Analysis |

nesc0539 | STRAP-2, Stress Analysis of Structure with Static Loading by Finite Elements Method |

nesc0539 | STRAP-D, Stress Analysis of Structure with Time-Dependent Loading by Finite Elements Method |

nea-0349 | STRESSPLOT, CALCOMP Plot of 2-D Finite Elements Calculation |

iaea0943 | STRIMP, Impurity Evolution in Tokamak Fusion Reactor Discharge |

nea-0253 | STYLE, Steam Cycle Heat Balance for Turbine Blade Design in Marine Operation |

nesc0924 | SUBDOSA, External Gamma, Beta Doses from Radionuclide Release into Atmosphere |

iaea1176 | SULSA, New Method for Neutron Spectrum Unfolding Problem |

nesc0056 | SUMMIT, Energy Transfer Diffusion Cross-Sections, Crystalline Moderator, Phonon Expansion |

nesc0638 | SUMUP-4, ENDF/B Utility, Partial Cross-Sections Sum Check Against Tot Cross-Sections |

psr-0282 | SUPERDAN-PC, Dancoff Factor for Spherical, Cylindrical, Slab Geometry |

iaea1437 | SUPERMC, Super Monte Carlo simulation program for nuclear and radiation process |

psr-0013 | SUPERTOG, MultiGroup Cross-Sections Generator from ENDF/B for Programs GAM, ANISN, DOT |

iaea0894 | SUPERTOG-LTT, SUPERTOG with Tabular Elastic Scattering Anisotropy from ENDL |

nesc9608 | SUPES, Engineering Sciences Utilities Program Library |

nesc0731 | SUPORT, Solution of Linear 2 Point Boundary Value Problems, Runge-Kutta-Fehlberg Method |

nesc0853 | SURGTANK, Steam Pressure, Saturation Temperature or Reactor Surge Tank |

nea-1151 | SUSD, Sensitivity and Uncertainty in Neutron Transport and Detector Response |

nea-1628 | SUSD3D, 1-, 2-, 3-Dimensional Cross Section Sensitivity and Uncertainty Code |

ccc-0248 | SWAN-PPL, Fusion Reactor 1-D Particle Transport Optimization |

ccc-0204 | SWANLAKE, Cross-Sections Sensitivity Analysis for 1-D Discrete Ordinate Calculation |

nesc0828 | SWAP-9, 1-D Stress Analysis for Hydrostatic and Elastic Plastic Materials |

nea-1698 | SWAT, Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2 |

nesc9811 | SWENT, 3-D Fluid, Heat, Radionuclide Transport in Heterogeneous Geologic Medium |

nesc0973 | SWIFT, 3-D Fluid Flow, Heat Transfer, Decay Chain Transport in Geological Media |

ests0682 | SWIMS, Sigmund and Winterbon Multiple Scattering of Ion Beams |

ccc-0767 | SWORD 6.0, SoftWare for Optimization of Radiation Detectors |

nesc0713 | SYN-3D, 2-D and 3-D Neutron Diffusion Static Eigenvalues, Single Channel Spatial Flux Synthesis |

nea-0594 | SYNTH-C, Steady-State and Time-Dependent 3-D Neutron Diffusion with Thermohydraulic Feedback |

iaea1383 | SYRCO-1, 1-D, 2 Groups Multi-Zone Interactive Diffusion Code |

nea-1023 | SYVAC, Risk Assessment from Underground Radioactive Waste Disposal in UK |

nesc9766 | T-HEMP3D, 3-D Time-Dependent Elastic Plastic Flow |

ests0219 | T2VOC, H2O, Air, VOC Flow Simulation in Porous Multidimensional Media |

nesc0408 | TAC-2D, Steady-State and Transient Heat Transfer in X-Y, R-Z or R-Theta Geometry |

nesc0414 | TAC-3D, 3-D Steady-State and Transient Heat Transfer in X-Y-Z and R-Theta-Z Geometry |

iaea0872 | TACHY, BWR Fuel Management by 2-D Coarse Mesh Neutron Diffusion |

nesc1113 | TACT-5, Doses of Radioactivity Release from Reactor Core into Environment |

nea-0532 | TAFE, 2-D Steady-State Heat Conduction for Structure with Gas Gaps |

nea-0531 | TAFEST, 2-D Transient Heat Conduction |

psr-0308 | TAM3, Monte-Carlo Sensitivity and Uncertainty Analysis of Radium in Lake Contamination Model |

nesc9566 | TAP-LOOP, Steady-State and Transient Thermal Analysis of Closed Test Loops |

nea-1301 | TAPE, General Copy Utility for VAX/VMS and IBM Tapes |

nea-0556 | TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements |

ccc-0638 | TART2016, 3D Coupled Neutron-Photon Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code |

nesc0558 | TASK, 1-D MultiGroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron |

ccc-0180 | TDA, Time-Dependent 1-D Neutron Transport, Gamma Transport by ANISN Method in Slab, Spherical, Cylindrical Geometry |

ccc-0256 | TDT, Time-Dependent and Steady-State Reactor Kinetics with Arbitrary Delayed Neutron Group |

ccc-0709 | TDTORT: Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System with Delayed Neutrons |

nesc9652 | TEKLIB, TEKTRONIX Graphics Subroutine Library |

nesc1084 | TEMAC, Top Event Sensitivity Analysis |

nea-0570 | TEMP, Steady-State and Transient Heat Conduction in Planar or Cylindrical Geometry |

iaea0836 | TEMPELS, Heat Conduction for Arbitrary Geometry by Finite Element Method (FEM) |

nesc0050 | TEMPEST-2, Thermalization Program for Neutron Spectra and MultiGroup Cross-Sections |

nesc9808 | TEMPEST-BNW, Transient 3-D Thermohydraulics for FBR |

nesc9653 | TEMPS, 1-Group Time-Dependent Pulsed Source Neutron Transport |

iaea1338 | TEMPUL, Temperature Distribution in Fuel Element after Pulse |

nea-1112 | TENDANCES, Search for Tendencies by Least Squares Fit Method |

nea-1328 | TERFOC-N, Radiation Doses in Food Chain from Atmospheric Release |

iaea1272 | THACT-RR, Analysis of Thermal Hydraulics Transients in Research Reactor Core |

nea-0774 | THALES, Thermohydraulic LOCA Analysis of BWR and PWR |

nea-1098 | THARC-S, Rod Bundle Thermohydraulic Transients of LMFBR for Single Phase Conditions |

nea-0634 | THERLIB, Library Generated for THERMOS from FACEL Library |

nesc9940 | THERMIT, 3-D Thermo-Hydraulics of BWR and PWR |

nea-0634 | THERMLIB, Generator and Edit of Program THERMOS-OTA Library |

nea-0043 | THERMOL, Space-Dependent Thermal Flux in 1-D Slab or Cylinder |

nesc0184 | THERMOS-ANL&BRT-1, 1-D Integral Transport for Neutron Spectra, Slab and Cylinder |

nea-0628 | THERMOS-OTA, Thermal Flux by Integral Transport |

nea-0411 | THESEE-3, Orgel Reactor Performance and Statistic Hot Channel Factors |

nesc0512 | THETA-1B, Fuel Rod Temperature Distribution by 2-D Diffusion, Heat Transfer to Coolant, LWR LOCA |

nea-0869 | THIDA, Transmutation, Hazard Potential, Dose Rate in Fusion Reactor |

nea-0869 | THIDA-2, Transmutation, Activation, Decay Heat, Dose Rate in Fusion Reactor |

nea-0377 | THREAT, 3-D Steady-State or Transient Heat Diffusion in Multi-Region Prism |

nesc0504 | THRES-2, Nuclear Induced Particle Emission Cross-Sections from Statistical Models |

nea-0658 | THRUSH, Thermal Neutron Coherent and Incoherent Scattering Kernels by Phonon Expansion |

nea-0997 | THT, 3-D Coarse Mesh LWR Bundle Fluxes and Power with Discontinuity Factors |

nea-0778 | THYDE-B2, Thermohydraulic Transients During LOCA of BWR |

nea-0779 | THYDE-P, PWR LOCA Thermohydraulic Transient Analysis |

nea-1592 | TIBSO, Nuclear Transitions and Radioactivity Migration in Technological System |

ests0643 | TIDY6.21, Reformatting of FORTRAN Source Programs |

nea-1077 | TIME-2, Radioactive Waste Disposal Climatic Change Risk Assessment |

nesc0756 | TIMEX, 1-D Time-Dependent MultiGroup Transport Theory with Delayed Neutron, Planar Cylindrical and Spherical Geometry |

nea-0387 | TIMOC-72, 3-D Time-Dependent Homogeneous or Inhomogeneous Neutron Transport by Monte-Carlo |

nea-0619 | TIMOC-ESP, Time-Dependent Response Function by Monte-Carlo with Interface to Program TIMOC-72 |

nea-0804 | TIMS-1, MultiGroup Cross-Sections of Heavy Isotope Mixture with Resonance from ENDF/B |

nea-0701 | TIRION-4, Atmospheric Dispersion of Radioactive Materials for Various Weather Conditions |

ccc-0759 | TITAN 1.29, A Three-Dimensional Deterministic Radiation Transport Code System |

ests0551 | TMAP4, Tritium Migration Analysis Program Version 4 |

ests0219 | TMVOCV1.0, Multicomponent, multiphase, nonisothermal flows of water, soil gas, volatile organic chemicals (VOCs) |

psr-0298 | TNG1, Multistep Statistical Model Hauser-Feshbach |

nesc9863 | TOEPLITZ, Solution of Linear Equation System with Toeplitz or Circulant Matrix |

nesc0561 | TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor |

nesc0561 | TOKMINA-2, Total Power for Tokamak Fusion Reactor |

nesc0627 | TOODY-2, Lagrangian Nonlinear Wave Propagation in 2-D X-Y or Cylindrical Geometry |

nesc1056 | TOOLPACK1, Tools for Development and Maintenance of FORTRAN 77 Program |

nesc9669 | TOPAZ-SNLL, Transient 1-D Pipe Flow Analysis |

iaea0909 | TOPIC-RUM, Plasma Impurities in Tokamak Reactor by MHD Method |

nea-1406 | TOPICS-B, Neutron and Gamma Cross-Sections Library Handling in FIDO Format |

nesc0599 | TOPLYR-2, Open Channel H2O Flow Temperature, Distant Source, Time-Dependent Boundary Conditions |

nesc1093 | TORAC, Flows, Pressure, Materials Transport within Structure During Tornado |

ccc-0543 | TORT, 2-D 3-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration |

nea-0486 | TOTEM, Demand Assessment for Nuclear Power Plants and Conventional Power Plants |

nesc1098 | TOUGH, Unsaturated Groundwater Transport and Heat Transport Simulation |

ests0219 | TOUGH2, Unsaturated Ground Water and Heat Transfer |

ests0219 | TOUGHREACTV1.2, Chemically reactive non-isothermal flows of multiphase fluids in porous and fractured media |

nesc9710 | TOXRISK, Toxic Gas Release Accident Analysis |

nea-1024 | TP2, Calculation of Reactivity and Kinetic Parameter by 2-D Neutron Transport. Perturbation Theory |

nea-0900 | TPHEX, MultiGroup Neutron Flux in Homogeneous Hexagonal LWR Cells |

nea-1070 | TPLOT, Interactive Postprocessor of Transient Structure Problems |

nea-1155 | TPTRIA, Reactivity for 2-D Triangular Geometry by Transport Perturbation Theory |

nesc0836 | TRAC, Thermohydraulics, Reactor Kinetics, 2 Phase Flow LOCA Analysis |

nesc1031 | TRAC-BD1, LOCA Analysis of BWR with 3-D Pressure Vessel and Multi Bundle Fuel Model |

nea-1593 | TRAC-PF1/EN MOD 3, Best Estimate Coupled 3-D Neutronics-Thermalhydraulics |

nea-1291 | TRANS-ACE, Radioactive Materials Transport in Reprocessing Plant Fire Accident |

nesc0268 | TRANS-FUGUE-1, Single Channel 2 Phase Flow Heat Transfer after Boiling |

nea-0953 | TRANSHEX, 2-D Thermal Neutron Flux Distribution from EpiThermal Flux in Hexagonal Geometry |

nesc0791 | TRANSPORT, Charged Particle Beam Transport 1st Order and 2nd Order Optical Analysis |

iaea1209 | TRANSV2, LOCA and Steady-State Thermohydraulic Analysis of MTR |

psr-0317 | TRANSX-2.15, Neutron Gamma Particle Transport Tables from MATXS Format Cross-Sections |

nea-0745 | TRAPSCO-2, Pressure and Temperature Transients in PWR Subcompartments During LOCA |

nea-0807 | TRAWA, LWR Dynamic by Coupled Neutron Diffusion and Thermohydraulics Calculation |

nea-0117 | TRAWS-4, Axial Flux Distribution for Control Rod Variations |

iaea0942 | TRAX, Resolution Matrix of Slow Neutron Spectrometers |

nea-0668 | TRD-3, In-Core and Out-Core Neutron Flux, Gamma Flux by 2-D Removal Diffusion in Cylindrical Geometry |

nesc1021 | TREDRA, Minimal Cut Sets Fault Tree Plot Program |

nea-0361 | TRESS, Triangular Mesh Stress and Strain in R-Z, X-Y Geometry for Various Load and Temperature |

ccc-0293 | TRIDENT, 2-D Neutron Transport for Homogeneous and Inhomogeneous Problems in X-Z, R-Z Geometry, Anisotropic Scattering |

iaea1214 | TRIGAC, Flux and Power Distribution and Burnup for TRIGA Reactor |

iaea1370 | TRIGLAV, Research Reactor Calculations |

nea-0384 | TRIGON, 2-D Homogeneous and Inhomogeneous Fixed Source Neutron Diffusion for Triangular or Hexagonal Mesh |

nesc1028 | TRIPM, Isothermal Transport and Decay of Radionuclides in Aquifer |

nea-1716 | TRIPOLI-4 version 8.1, 3D general purpose continuous energy Monte Carlo Transport code |

nea-1878 | TRIPOLI-4 version 9S, Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo Transport Calculation |

ccc-0537 | TRIPOS, Monte Carlo Ion Transport Analysis Code |

nea-1086 | TRISTAN, 3-D fixed source radiation transport |

iaea1337 | TRISTAN-IJS, Steady-State Axial Temperature and Flow Velocity in Triga Channel |

nea-1087 | TRITAC, 3-D Transport by Discrete Ordinate Method in X-Y-Z Geometry |

ests0308 | TRITMOD, Environmental Transport and Cycling of H3 after Atmospheric Releases |

nea-0415 | TRITON, 3-D Multi-Region Neutron Diffusion Burnup with Criticality Search |

iaea0884 | TRIVENI, 3-D Fuel Management for PHWR CANDU |

psr-0522 | TRUMP, Steady-State and Transient 1-D, 2-D and 3-D Potential Flow, Temperature Distribution |

nea-0233 | TURBINA, Reheat Steam Turbine Generator Design with Preheater and Condenser |

nea-0581 | TURBPLANT, 1-D Steady-State Model of Power Reactor Steam Turbine Components |

nesc0042 | TUZ, Resonance Integrals in Unresolved Region, Various Temperature, From Porter-Thomas Distribution |

nea-0471 | TVEDIM, 2-D Homogeneous and Inhomogeneous Neutron Diffusion for X-Y, R-Z, R-Theta Geometry |

nesc0809 | TVENT, 1-D Incompressible Flow for Pressure Transients in Ventilation System |

nesc0712 | TWODEE-2/MOD3, 2-D Time-Dependent Fuel Elements Thermal Analysis after PWR LOCA |

nesc0358 | TWOTRAN-2, 2-D MultiGroup Transport in X-Y, R-Z, R-Theta Geometry with Anisotropic Scattering |

ccc-0195 | TWOTRAN-GG-FC, General Geometry 2-D Transport with 1st Collision Source Calculation |

ccc-0195 | TWOTRAN-GG-VW, General Geometry 2-D Transport, Variable-Weight Diamond Difference |

iaea1434 | U-SHIELDER, Estimates Shielding Thickness of Depleted Uranium for Photons from 0.5 to 10 MeV |

nesc9668 | UCBNE, Radionuclide Migration in Porous Media |

nesc9667 | UCBNE25, Radionuclide Migration in Geologic Media |

nesc0824 | UDAD, Radiation Exposure to Man at Uranium Processing Plant |

ests0404 | UHS, Ultimate Heat Sink Cooling Pond Analysis |

psr-0015 | UKE, Format Conversion from UKNDL to ENDF/B |

nea-1665 | UMG 3.3, Analysis of data measured with spectrometers using unfolding techniques |

nea-1139 | UNC32/33, Covariance Matrices from ENDF/B-5 Resonance Parameter Uncertainties |

nea-0175 | UNCLE, Crystal Scattering Kernel with Coherent Scattering by Butler Approximation |

iaea1242 | UNF, Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials |

iaea1177 | UNIFY, Fast Neutron Cross-Sections and Spectrum for Structural Materials |

ests0827 | UNSPEC, X-Ray Spectrum Unfolding |

iaea0959 | UPEAK, General Experimental Spectra Analysis Program |

psr-0245 | UPEML, Computer Independent Emulator of CDC Update Utility |

psr-0281 | URR, Cross-Sections, Selfshielding for Fertile and Fissile Isotopes in Unresolved Region |

ests0333 | USINT, High Temperature Heat and Mass Transfer on Concrete Surfaces in LMFBR |

nesc9848 | UTAH-2, Thermoplastic Response in Anisotropic Rock |

uscd1150 | UTAP, U Tailings Assessment Program |

ccc-0500 | UTMTOX, Toxic Chemical Transport in Atmosphere, Ground Water, Sediments |

nea-0356 | UTOE, UKNDL to ENDF/B Format Conversion with Log-Log Interpolation and Angular Distribution Tables |

nea-0587 | UTSG, Steady-State and Transients of Vertical U-Tube Steam Generator |

ccc-0613 | VALE-1.1, 2-D, 3-D MultiGroup Neutron Diffusion for Triangular Problems |

nesc0264 | VARI-QUIR-3, 2-D MultiGroup Steady-State Neutron Diffusion in X-Y R-Z or R-Theta Geometry |

nesc0755 | VARR2 VARRLXSG, 2-D Transient Fluid Flow and Heat Transfer in X-Y and Cylindrical Geometry |

ccc-0781 | VARSKIN 4 V4.0.0, Dose Calculation for Skin Contamination, with Sadde Input Generator |

ests0752 | VCODE, Ordinary Differential Equation Solver for Stiff and Non-Stiff Problems |

ccc-0262 | VCS, Radiation Protection Factors in Vehicles by Monte-Carlo |

uscd1239 | VENTEASY, Criticality Search for a Desired Keffective by Adjusting Dimensions, Nuclide Concentrations, or Buckling |

ccc-0654 | VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup |

nesc0511 | VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions |

nesc9826 | VERTPAK-1, Fluid Flow, Rock Deformation, Solute Transport in Porous Media |

nea-1856 | VESTA 2.1.5, Monte Carlo depletion interface code and AURORA 1.0.0, Depletion analysis tool |

psr-0311 | VIDEO-PC, SVGA Routines for FORTRAN on PC |

ccc-0754 | VIM 5.1, Steady-State 3-D Neutron Transport Using ENDF/B or Multigroup Cross Sections |

nesc0510 | VIM, 3-D Monte-Carlo Analysis of Fast Critical Assemblies Using Point Cross-Sections |

uscd1240 | VIM_NC, VIM color syntax for Nuclear Codes: NJOY, DRAGON, PARTISN, TORT, MONK, and MCNP |

iaea0932 | VIRGIN2010, Calculates Uncollided Neutron Flux and Neutron Reactions from Transmission in ENDF Format |

nesc1115 | VISA-2, Reactor Vessel Failure Probability Under Thermal Shock |

nesc9846 | VISCOT, Viscous Mechanical Behaviour of Rock Mass Under Thermal Stress |

psr-0618 | VISUAL EDITOR 61, MCNPX/6 Visual Editor Computer Code |

iaea1324 | VITEK, Non Stationary Navier-Stokes Solver for Compressible, Turbulent Flow |

nea-0636 | VIWI, Neutron Speeds and Weights for Scattering Kernel Calculation |

nesc0922 | VMCON, Minimization of Nonlinear Function with Constraints |

ests0426 | VODE, Variable Coefficient Ordinary Differential Equations (ODE) Solver |

iaea0871 | VPI-NECM, Nuclear Engineering Program Collection for College Training |

nea-0655 | VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation |

iaea1417 | W-SHIELDER, calculates shielding thickness of water for photon emitting radionuclide between 0.5 to 10 MeV |

nea-1142 | WADOSE, Radiation Source in Vitrification Waste Storage Apparatus |

nea-0506 | WAKE, Navier Stokes Equation with 2-D Turbulence, Stream Function, Vorticity |

nesc9673 | WAPPA, Waste Package Performance Assessment |

uscd1157 | WATEQ4F, Aqueous Speciation Calculation of Natural Waters |

iaea1210 | WEDRO, Data Processing Routines for WIMS-D/4 WIMSE File |

nea-0610 | WEERIE, Radioactive Release from Reactor to Cooling Circuit and Atmosphere |

ests0160 | WELBORE, Transient Wellbore Fluid Flow Model |

ests1197 | WFSFIT, Wilson-Fowler Spline Fit Algorithm |

nesc0278 | WHAM-6, Pressure and Velocity Transients in Fluid Pipes, Wave Superposition Method |

nea-1147 | WHATIF-AQ, Geochem Speciation and Saturation of Aqueous Solution |

iaea1243 | WILIT, Utility Program for WIMS Library Handling |

iaea0946 | WILMA, WIMS Nuclear Data Library Maintenance |

nea-0329 | WIMS, MultiGroup Reactor Lattice Calculation for Thermal Reactor and Fast Reactor |

ccc-0698 | WIMS-ANL 4.0, Deterministic Code System for Lattice Calculation |

iaea0887 | WIMSCORE, 2-Group Constant from WIMS-D/4 for Programs TDB, TRITON, CITATION |

nea-1507 | WIMSD5, Deterministic Multigroup Reactor Lattice Calculations |

iaea1254 | WINTER, Interactive WIMS Input Preparation |

iaea1408 | WLUP3.0, 69 and 172 Group Cross Section Libraries for WIMS |

ccc-0427 | WRAITH, Internal and External Doses from Atmospheric Release of Isotopes |

iaea0897 | X4ECS, ENDF/B-4 and EXFOR Data Comparison |

iaea0896 | X4R, EXFOR Evaluated Data Retrieval |

iaea0936 | X4TOC4, Neutron Cross-Sections Data Conversion from EXFOR to Computation Format |

nea-0564 | XBWR, 1-D Xe Transients for BWR in Axial Geometry |

nesc0988 | XERROR, FORTRAN Library Error Message Processing Routines |

nesc0572 | XLACS, Fast Resonance and Thermal MultiGroup Cross-Sections from ENDF/B, Breit-Wigner, for Program XSDRN |

iaea1395 | XNWLUP, Graphical user interface to plot WIMS-D library multigroup cross sections |

nesc0964 | XOQDOQ, Meteorological Evaluation of Atmospheric Nuclear Power Plant Effluents |

ccc-0525 | XRAY-AAC, X-Ray Attenuation and Absorption |

nesc0393 | XSDRN, MultiGroup Cross-Sections from Resonance Data Library, Neutron Spectra and Group Constant Collapsing |

nea-1882 | XSUN-2017, Windows interface environment for transport and sensitivity-uncertainty software TRANSX-2, PARTISN and SUSD3D |

nea-0072 | ZADOC, 2 Group Time-Dependent Burnup in X-Y Geometry with Fuel Management |

nea-0283 | ZEUS-ALB.5, 3-D 1 Group Neutron Transport Kinetics in Slits, Channels, Tunnels by Monte-Carlo |

nea-0401 | ZOCO-6, Temperature Transients in BWR and PWR Containment During LOCA |

iaea1371 | ZOTT99, Data Evaluation Using Partitioned Least-Squares |

nesc0041 | ZUT, Resonance Integrals in Resolved Region at Various Temperature, Escape Probability Calculation |

nea-1251 | ZYLIND, Gamma Penetration for Cylindrical Source and Shield Geometry |

nea-0789 | ZZ ABBN, 26 Group Cross-Sections and Self Shielding Factors for Fast Reactors |

nea-0790 | ZZ ACTINIDES, 84-Group Neutron Cross-Section Library for Pu242 to Es253 Isotope Production Chain |

iaea1275 | ZZ ACTIV-87, Fast Neutron Activation Cross-Section |

dlc-0069 | ZZ ACTL82, Data Library of Evaluated Activation Cross-Sections |

iaea1420 | ZZ ADS-LIB/V1.0, test library for Accelerator Driven Systems v.1.0 |

iaea1420 | ZZ ADS-LIB/V2.0, test library for Accelerator Driven Systems v.2.0 |

dlc-0014 | ZZ AIR, Group Constant Library of Secondary Gamma Transport in Air for ANISN Calculation |

dlc-0049 | ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section |

dlc-0224 | ZZ ALBEDO-DATA, Data for the Calculation of Albedos from Concrete, Iron, Lead and Water for Photons and Neutrons |

nea-1745 | ZZ ALEPH-LIB-JEFF3.1, MCNP Neutron Cross Section Library based on JEFF3.1 |

nea-0886 | ZZ AMPX-2/123, 123-Group Neutron Cross-Section Library from ENDF/B-4 by AMPX-2 |

nea-0886 | ZZ AMPX-2/219, 219-Group Neutron Cross-Section Library from ENDF/B-4 by AMPX-2 |

dlc-0027 | ZZ AMPX01/27C, Coupled Neutron-Gamma Group Constant Library by AMPX for Transport Calculation |

iaea0912 | ZZ AMZ, 70-Group 40 Isotope Multigroup Library for Fast Reactor Calculation |

dlc-0129 | ZZ ANS643, Geometric Progression Gamma-Ray Buildup Factor Coefficient Library |

dlc-0154 | ZZ ANSLV, Multigroup Cross Sections Library for ANS Reactor Design Studies |

iaea1278 | ZZ ASIYAD, Fission-Product Yield Data Library for Neutron-Induced Fission |

nea-0673 | ZZ BABEL, Multigroup Neutron Cross-Section Data Library for Fast Reactor Shield Calculation |

iaea0856 | ZZ BARC-27GRP, 27-Group Infinitely Dilute and Bondarenko Cross-Section Library from ENDF/B |

iaea1237 | ZZ BARC-75, Coupled 50 Neutron-Group 25 Gamma-Group Cross-Section Library for 42 Nuclides |

iaea1398 | ZZ BOREHOLE-EB6.8-MG, multi group cross-section library for deterministic and Monte Carlo codes |

dlc-0008 | ZZ BP-3, 104-Group Neutron Cross-Section Library for Transport Calculation |

dlc-0008 | ZZ BP-6, 104 Group Neutron and Gamma-Ray Multigroup Cross-Section Library for Transport Calculation |

iaea0949 | ZZ BROND, Evaluated Neutron Data Library in ENDF-6 Format |

nea-1872 | ZZ BUGENDF70.BOLIB, ENDF/B-VII.0 Broad-Group Coupled X Sect. Lib. for LWR Shielding & Pressure Vessel Dosimetry Applic. |

nea-1866 | ZZ BUGJEFF311.BOLIB, JEFF-3.1.1 Broad-Group Coupled X Sect Lib. for LWR Shielding & Pressure Vessel Dosimetry Applic. |

dlc-0185 | ZZ BUGLE-96, Multigroup Coupled Neutron Gamma Cross-Section for LWR Shielding Calculation |

dlc-0059 | ZZ CAD, 51 Neutron-Group, 25 Gamma-Group Albedo Data for 4 Materials from DOT Flux |

dlc-0210 | ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for CANDU Reactor Fuels |

iaea1259 | ZZ CENDL, Evaluated Nuclear Data Library for Neutron Reaction Data |

iaea1256 | ZZ CENPL, Chinese Evaluated Nuclear Parameter Library |

iaea1256 | ZZ CENPL-DLS, Discrete Level Schemes and Gamma Branching Ratios Library |

iaea1256 | ZZ CENPL-FBP, Fission Barrier Paramater Library |

iaea1256 | ZZ CENPL-GDRP, Giant Dispole Resonance Parameter Library |

iaea1256 | ZZ CENPL-MCC, Nuclear Ground State Atomic Masses Library |

iaea1256 | ZZ CENPL-NLD, Nuclear Level Density Parameter Library |

iaea1256 | ZZ CENPL-OMP, Optical Model Parameter Library |

iaea1297 | ZZ CL50G, 50-Group Multigroup Library in AMPX Format for Fast Reactor Calculation |

dlc-0042 | ZZ CLEAR/42B, 126 Neutron-Group, 36 Gamma-Group Coupled Cross-Section in AMPX, CCCC Format, for LMFBR |

nea-1775 | ZZ CLES, cross section library of moderator materials for low-energy neutron sources |

dlc-0016 | ZZ COBB, 123 Neutron-Group Cross-Section Library from ENDF/B for XSDRN Calculation |

dlc-0271 | ZZ COG-SUP-LIB, COG Supplemental Libraries for ENDL2011 and MCNP6.1-ENDF/B-VII. 1 |

nea-1730 | ZZ COV-15GROUP-2006, 15-group cross section covariance matrix library |

dlc-0077 | ZZ COVERV, Multigroup Cross-Section Covariance Matrices in COVERX Format |

dlc-0091 | ZZ COVFILS, 30-Group Covariance Library from ENDF/B-5 for Sensitivity Studies |

dlc-0137 | ZZ COVFILS-2, 74-Group Neutron Cross-Section, Scattering Matrices, Covariances for Fusion Reactors |

dlc-0138 | ZZ COVFILS-2-I, 74-Group Neutron Cross-Section, Scattering Matrices, Covariances for Fusion Reactors |

nea-1787 | ZZ CRYO-S(A,B)-ACE1, Scattering law and continuous energy cross section library of materials at cryogenic temperatures |

dlc-0028 | ZZ CTR, 73-Group Neutron and Gamma Coupled Cross-Section for CTR Transport Calculation |

dlc-0130 | ZZ DABL69, 46-Group Neutron, 23-Group Gamma Cross-Section in ANISN Format from ENDF/B-V |

nea-0791 | ZZ DAMSIG84, 640-Group Damage Cross-Section Library for SAND-2 Calculation |

dlc-0030 | ZZ DECAYREM/C, Decay Spectra Library for EXREM Calculation |

nea-1644 | ZZ DECDC, Nuclear Decay Data Files for Dose Calculation |

nea-1538 | ZZ DECNET-GENDF, Fusion Damage Library of 175 Neutron and 42 Photon VITAMIN-J Groups |

dlc-0010 | ZZ DLC-10B, Neutron Kerma Response Function Data Library |

dlc-0011 | ZZ DLC-11/RITTS, 121-Group Coupled Cross-Section for ANISN, DOT, MORSE |

dlc-0019 | ZZ DLC-19 DECAYGAM, Isotope Gamma Energy Library for Spectrometry Evaluation |

dlc-0023 | ZZ DLC-23F CASK, 40-Group Neutron and Gamma Coupled Cross-Section for PWR Shipping Casks |

dlc-0002 | ZZ DLC-2D/100G, 100 Neutron-Group Cross-Section Library by SUPERTOG Calculation for ANISN, DOT |

dlc-0031 | ZZ DLC-31, 37 Neutron-Group, 21 Gamma-Group Coupled Group Constants Library from ENDF/B |

dlc-0090 | ZZ DOSCOV, 24-Group Covariance Data Library from ENDF/B-V for Dosimetry Calculation |

nea-0827 | ZZ DOSCROS84, Neutron Cross-Section Library for Spectra Unfolding and Integral Parameter Evaluation |

dlc-0079 | ZZ DOSDAT-2, Gamma and Electron Dose Conversion Factor Data Library for Body Organs |

dlc-0144 | ZZ DOSEDAT-DOE, Dose-Rate Conversion Factors for External Photon, Electron Exposure |

dlc-0080 | ZZ DRALIST, Radioactive Decay Data for Dosimetry and Hazard Assessment |

iaea1401 | ZZ DROSG-2000, Legendre Coefficient Library for 59 monoenergetic neutron source reactions |

nea-1609 | ZZ EAF 99, Cross Section Library for Neutron Induced Activation Materials |

dlc-0106 | ZZ ECPL86, Data Library of Evaluated Charged Particle Cross-Section, Nuclides Up to Oxygen |

nea-1050 | ZZ EFF1LIB, Fusion Fast Neutron Data Library for MCNP |

dlc-0208 | ZZ ELAST2, Database of Cross Sections for the Elastic Scattering of Electrons and Positrons by Atoms |

dlc-0100 | ZZ ELECSPEC, Electron Spectra Data Library from Fission Product Decay |

uscd0803 | ZZ ENDF/B-IV, Evaluated Nuclear Data File Version 4 |

uscd1233 | ZZ ENDF/B-V, Evaluated Nuclear Data File Version 5 |

dlc-0103 | ZZ ENDL86, Evaluated Charged Particle, Neutron, Photon Cross-Section Library |

dlc-0179 | ZZ ENDLIB, Coupled Electron and Photon Transport Library in ENDL Format |

iaea1435 | ZZ EPICS2017, Electron Photon Interaction Cross Sections |

dlc-0037 | ZZ EPR/37F, 100 Neutron-Group, 21 Gamma-Group Coupled Cross-Section for Experimental Power Reactor (EPR) Fusion System |

nea-0794 | ZZ EURLIB, Coupled Neutron Gamma Multigroup Cross-Section Library from ENDF/B for Shielding Calculations |

dlc-0085 | ZZ FCXSEC, Coupled Cross-Section Library for Shielding from VITAMIN-C in AMPX, ANISN Format |

dlc-0167 | ZZ FGR-DOSE, Dose Coefficient for Intake and Exposure to Radionuclides |

iaea0964 | ZZ FGXRRS, 10 Neutron-Group, 7 Gamma-Group Self-Shielded Cross-Section in ANISN Format |

nea-1822 | ZZ FLUKA05-PRE-LIB, FLUKA05 Multi-group, multi-purpose nuclear data library, neutrons, photons, charged particles |

nea-1424 | ZZ FSXJ32, MCNP nuclear data library based on JENDL-3.2 |

nea-1782 | ZZ FSXLIB-JD99, MCNP nuclear data library based on JENDL Dosimetry File 99 |

nea-1424 | ZZ FSXLIBJ33, MCNP nuclear data library based on JENDL-3.3 |

nea-0878 | ZZ GAMDAT-78, Gamma Decay Data of Radioisotopes |

dlc-0006 | ZZ GAMLIB, 99-Group Cross-Section Library by SUPERTOG Calculation from ENDF/B |

dlc-0071 | ZZ GAMMON, Activation Data Library for Fusion Reaction |

dlc-0013 | ZZ GARLIB, Multigroup Resonance Cross-Section Group Constant Library for Tungsten and Depleted Pu |

nea-1543 | ZZ GEFF-2-GENDF, P5 175-N and P8 42-Gamma Group Library for Fusion Blanket Applications |

nea-1544 | ZZ GEFF-2-MATXS, Coupled Neutron-Gamma Fusion Neutronics Library in MATXS Format |

nea-1102 | ZZ GEFF1, 175-Group Neutron Cross-Section in VITAMIN-J1 Format for Shielding Benchmarks |

nea-1344 | ZZ GROUPSTRUCTURES, VITAMIN-J, XMAS, ECCO-33, ECCO2000 Standard Group Structures |

nea-1210 | ZZ HATCHES-20, Database for radiochemical modelling |

dlc-0220 | ZZ HILO2K, Coupled 83 Neutron, 22 Photon Group Cross Sections for Neutron Energies Up to 2 GeV for ANISN, DORT and TORT |

dlc-0119 | ZZ HILO86, 66 Neutron, 22 Gamma Group Cross-Section Library for ANISN, DORT, MORSE |

dlc-0187 | ZZ HILO86R, 66 Neutron, 22 Gamma Group Cross-Section for 400 MeV Neutron, 20 MeV Gamma |

dlc-0007 | ZZ HPICE, Gamma Interaction Cross-Section Library in ENDF/B Format for Transport Calculation |

dlc-0099 | ZZ HUGO, Photon Interaction Data Library in ENDF-5 Format |

dlc-0146 | ZZ HUGO-VI, Photon Interaction Data in ENDF-6 Format |

iaea1419 | ZZ IBANDL, Ion Beam Analysis Nuclear Data Library in R33 format |

nea-1656 | ZZ IEAF-2001, Intermediate Energy Activation File |

iaea1418 | ZZ INDL/TSL, Thermal Neutron Scattering Data for H2O, D2O and ZrHx in ENDF-6 Format and as MCNP(X) Data Sets |

iaea1215 | ZZ IRAN-LIB, Multigroup Neutron Gamma Cross-Section Library for 33 Elements in ANISN Format |

iaea0867 | ZZ IRDF-2002, 640-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-6 Format |

iaea0867 | ZZ IRDF-2002-ACE, Cross-Section Library and Spectra for Dosimetry Calculation in ACE Format for Monte Carlo methods |

iaea0867 | ZZ IRDF-82, 620-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-5 Format |

iaea0867 | ZZ IRDF-90, 640-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-6 Format |

nea-1853 | ZZ JENDL-1, Japanese Evaluated Nuclear Data Library |

nea-1624 | ZZ JENDL/D-99, JENDL Dosimetry Cross-Sections Data Library and Graphical Representations |

nea-0796 | ZZ JFS-1, Cross-Sections Library 25-Groups ABBN and 70-Group JFS for Fast Reactor Calculation |

nea-0796 | ZZ JFS-2, 25 Group (ABBN) and 70 Group JFS Cross Sections Library for Fast Reactors |

nea-0796 | ZZ JFS-3/J2, 70 Group 30 Isotopes Cross Section Library for Fast Reactors |

nea-1815 | ZZ KAFAX-E70, 150 and 12 Groups Cross Section Library in MATXS Format based on ENDF/B-VII.0 for Fast Reactors |

nea-1650 | ZZ KAFAX-F22, 80 and 24 Groups Cross-Section Library in MATXS Format Based on JEF-2.2 for Fast Reactors |

nea-1816 | ZZ KAFAX-F31, 150 and 12 Groups Cross Section Library in MATXS Format based on JEFF-3.1 for Fast Reactors |

nea-1817 | ZZ KAFAX-J33, 150 and 12 Groups Cross Section Library in MATXS Format based on JENDL-3.3 for Fast Reactors |

dlc-0160 | ZZ KAOS/LIB-V, Kerma Factors, Nuclear Response Function Library for Fission, Fusion |

nea-1649 | ZZ KASHIL-E6, 175 N, 42 Gamma Groups Cross Sections in MATXS Format Based on ENDF/B-VI.5 for Shielding Applications |

nea-1818 | ZZ KASHIL-E70, 199 N, 42 Photon Groups Cross Sections in MATXS Format Based on ENDF/B-VII.0 for Shielding Applications |

dlc-0142 | ZZ KERMAL, Neutron and Gamma Kerma Library from ENDL and EGDL |

dlc-0021 | ZZ KXRAY, X-Ray Attenuation Cross-Section Library from 0.1 KeV to 1 MeV |

iaea0870 | ZZ L26P3S34, 26-Group Constants Library of 34 Materials for Neutron Shielding Calculations |

dlc-0168 | ZZ LA100, ENDF Format Data Library for Neutron and Protons Up to 100 MeV |

dlc-0054 | ZZ LAFPX-V, Multigroup Fission Product Data Library from ENDF/B-V by Program NJOY |

dlc-0128 | ZZ LAHIMACK, Multigroup Neutron and Gamma Cross-Section and Response Function up to 800 MeV |

nesc0532 | ZZ LASL-XSECS, Fast and Thermal Multigroup Cross-Section Library in LANL Transport Format |

dlc-0040 | ZZ LIB-IV, 50-Group Cross-Section Library in CCCC-III Format from ENDF/B-IV for Fast Reactors |

dlc-0089 | ZZ LUMP, Lumped Fission Product Cross-Section Library for Fast Reactor Analysis from ENDF/B-V |

dlc-0029 | ZZ MACKLIB, Nuclear Response Function Library for CTR and Hybrid Fission Fusion System Materials |

dlc-0060 | ZZ MACKLIB-4, 171-Neutron, 36-Gamma Group Response Function Library from ENDF/B-IV |

nea-1740 | ZZ MATJEF22.BOLIB, JEF-2.2 Multigr Coupled (199n + 42gamma) X-Section Lib. in MATXS Fmt for Nuclear Fission Applications |

nea-1847 | ZZ MATJEFF31.BOLIB, JEFF-3.1 Multigr Coupled(199n + 42gamma) X-Section Lib.in MATXS Fmt for Nuclear Fission Applications |

nea-1205 | ZZ MATX175/42-JEFF87, 172 Neutron-Group, 42 Gamma-Group MATXS Library in VITAMIN-J Structure |

dlc-0176 | ZZ MATXS10, 30-Group Neutron, 12-Group Gamma Cross-Sections in MATXS Format from ENDF/B-VI |

dlc-0177 | ZZ MATXS11, 80-Group Neutron, 24-Group Gamma Cross-Section in MATXS Format from ENDF/B-VI |

nea-1206 | ZZ MATXS70-JEFF87, 69+1 Group MATXS Library in WIMS BOXER Structure |

nea-1707 | ZZ MATXSLIBJ33,JENDL-3.3 based,175 N-42 photon groups (VITAMIN-J) MATXS lib. for discrete ordinates multi-group |

nea-1668 | ZZ MCB-EAF99, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K. |

nea-1669 | ZZ MCB-ENDF/B6.8, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K. |

nea-1667 | ZZ MCB-JEF2.2, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K. |

nea-1670 | ZZ MCB-JENDL-3.2, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K. |

nea-1655 | ZZ MCB63NEA.BOLIB, MCNP Cross Section Library Based on ENDF/B-VI Release 3 |

nea-1616 | ZZ MCJEF22NEA.BOLIB, MCNP Cross Section Library Based on JEF-2.2 |

nea-1768 | ZZ MCJEFF3.1NEA, MCNP Neutron Cross Section Library based on JEFF3.1 |

nea-1651 | ZZ MCLIB-E6, Continuous Energy Cross Section Library from ENDF/B-VI.5 for MCNP-4A, -4B, 300K, 600K, 900K |

dlc-0200 | ZZ MCNPDATA, ZZ-MCB-DLC200, Standard Neutron, Photon and Electron Data Libraries for MCNP-4C and MCB1C |

iaea1376 | ZZ MENDL-2P, Proton Medium Energy Nuclear Data Library |

nea-1613 | ZZ MICROX 2 FSS LIB, Data Library for Fast Spectrum Systems Analysis |

iaea1412 | ZZ MINSKACT, Evaluated neutron reaction data for Th-232, Pa, U, Np, Pu, Am and Cm isotopes |

dlc-0033 | ZZ MONTAGE-400, Neutron Activation 100-Group Cross-Section Library of Fusion Reactor Materials |

iaea1217 | ZZ N-SPECT/DET-RESP, Neutron Spectra and Detector Responses for Radiation Protection |

dlc-0018 | ZZ NAB, 100 Neutron-Group Cross-Section Library of Na and Al for ANISN, DOT, MORSE Neutron Transport |

iaea1279 | ZZ NMF-90, Database for Neutron Spectra Unfolding |

dlc-0017 | ZZ NOX, 119-Group Coupled Cross-Section of Nitrogen, Oxygen, Air for MORSE |

dlc-0172 | ZZ NUCDECAY, Nuclear Decay Data for Radiation Dosimetry Calculation for ICRP and MIRD |

dlc-0202 | ZZ NUCDECAYCALC, Nuclear Decay Data for Radiation Dosimetry Calculation for ICRP |

nea-1642 | ZZ ORIGEN2.2-UPJ, A complete package of ORIGEN2 libraries based on JENDL-3.2 and JENDL-3.3. |

nea-1642 | ZZ ORLIBJ32, ORIGEN2 libraries based on JENDL-3.2 |

dlc-0038 | ZZ ORYX-E, Group Constant Library from ENDF/B Fission Product Data for ORIGEN Calculation |

iaea1423 | ZZ PADF-2007, Proton Activation Data File in ENDF-6 format |

dlc-0236 | ZZ PHOBIA, Photon Buildup Factors to Account for Angular Incidence on Shield Walls |

dlc-0136 | ZZ PHOTX, Photon Interaction Cross-Section Library for 100 Elements |

nea-1868 | ZZ PIXE2010, Proton/Alpha Ionization (K,L,M shell) Tabulated Cross-Section Library |

iaea1235 | ZZ PNESD, Diffusion Elastic Scattering Cross-Section of 3 MeV to 1000 MeV Proton on Natural Isotopes |

iaea1409 | ZZ POINT-2004, Linearly Interpolable ENDF/B-VI.8 Data for 13 Temperatures |

iaea1421 | ZZ POINT-2007, linearly interpolable ENDF/B-VII.0 data for 14 temperatures |

iaea1430 | ZZ POINT-2009, a Temperature Dependent ENDF/B-VII.0 Cross Section Library |

dlc-0212 | ZZ POINT2000, Linearly Interpolable ENDF/B-VI.7 Data for 8 Temperatures |

dlc-0247 | ZZ POINT2011, Linearly Interpolable ENDF/B-VII.1 Beta2 Cross-Section Library for 13 Temperatures |

dlc-0192 | ZZ POINT97, Temperature-Dependent ENDF/B-6 Cross-Sections at 8 Temperature Between 0K and 2100K |

dlc-0012 | ZZ POPLIB, Secondary Gamma Yields and Cross-Section Library for POPOP-4 Calculation |

dlc-0196 | ZZ PR-EDB, Power Reactor Embrittlement Database |

iaea1277 | ZZ PRONDOS, Evaluations of Selected Neutron Activation Reactions for Dosimetry |

dlc-0126 | ZZ PVE, 38-Group P8 Photon Cross-Section Library for Gamma Radiation Transport |

dlc-0134 | ZZ RADDECAY, Decay Data Library for Radiological Assessment |

nesc9554 | ZZ REAC-2, Nuclide Activation and Transmutation |

nea-1255 | ZZ REAC-ECN-3/GEAF-1-2/EAF-3, Neutron Reaction Cross-Sections Library for Fusion Reactors |

dlc-0055 | ZZ RECOIL/B, Heavy Charged Particle Recoil Spectra Library for Radiation Damage Calculation |

nea-1545 | ZZ RFL-2-DTF, Group Constant Library of Reaction Cross-Section, Gas Production, KERMA, DPA |

iaea1365 | ZZ RIPL, ZZ RIPL-2, Parameter Library for Nuclear Model Calculations |

iaea1250 | ZZ RNPL-A, Nuclear Masses Library |

iaea1407 | ZZ RRDF-98, Cross-sections and covariance matrices for 22 neutron induced dosimetry reactions |

dlc-0057 | ZZ SAIL, Albedo Scattering Data Library for 3-D Monte-Carlo Radiation Transport in LWR Pressure Vessel |

dlc-0076 | ZZ SAILOR, 47 Neutron-20 Gamma-Group Coupled Cross-Section Library from VITAMIN-C by AMPX |

nea-1185 | ZZ SCALE-LIB, Neutron-Group Constants Library from JEF-1 Using NPTXS, XLACS, XLACS-2 Programs |

uscd1236 | ZZ SCALE5.1/COVA-44G, 44-group cross section covariance matrix library extracted from SCALE5.1 |

uscd1236 | ZZ SCALE6.0/COVA-44G, 44-group cross section covariance matrix library extracted from SCALE6.0 |

dlc-0045 | ZZ SENPRO/45C, Multigroup Sensitivity Library for Fast Reactors, Thermal Reactors |

nea-1854 | ZZ SERPENT117-ACELIB, Continuous-energy X-sec lib., radioactive decay, fission yield data for SERPENT in ACE |

dlc-0135 | ZZ SHAMSI, Coupled 43-Neutron 14-Gamma P3 Cross-Section Library for Fusion Blanket or Shield Calculations |

dlc-0139 | ZZ SIGMA-A, Photon Interaction and Absorption Cross-Section Library |

dlc-0024 | ZZ SINEX, 100 Neutron-Group Neutron Reaction Cross-Section Library from ENDF/B by SUPERTOG for ANISN |

dlc-0188 | ZZ SKYDATA-KSU, Neutron and Gamma Skyshine Responses |

dlc-0093 | ZZ SKYPORT, Importance Function for Neutron and Gamma for Skyshine Dose from Accelerator |

dlc-0178 | ZZ SNLRML, Dosimetry Cross-Section Recommendations |

dlc-0015 | ZZ STORM-ISRAEL, Gamma Interaction Cross-Section Library in ENDF/B Format for Transport |

iaea0865 | ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD |

nea-1837 | ZZ TENDL-2008-ACE, TENDL-2008 based library for neutrons, protons, deuterons, tritons, helions, alpha and gammas |

iaea1399 | ZZ TH-N-CAPTURE-RI&G, Thermal Neutron Capture Cross Sections Resonance Integrals and G-Factors |

nea-1674 | ZZ TH232-UNIBO, Th-232 cross section data for MCNP |

dlc-0140 | ZZ THERMGAM, Thermal Neutron Capture Gamma Spectroscopical Data Library |

nesc0543 | ZZ THERMOS, Multigroup P0 to P5 Thermal Scattering Kernels from ENDF/B Scattering Law Data |

dlc-0088 | ZZ TPASGAM-85, Gamma Spectra Data Library for Activation Analysis |

nea-1883 | ZZ TSL-ACE/2013, Thermal Scattering Libraries processed to ACE format |

nea-0899 | ZZ UKCNDL-82, Chemical Nuclear Data Library of Fission and Decay Reactions in ENDF Format |

nea-0642 | ZZ UKCTR-1, Cross-Section Library for Neutron Flux and Neutron Reaction Rates in CTR Calculation |

nea-0680 | ZZ UKCTRIIIA, Neutron Cross-Section Data Library for Fusion Reactor Materials Activation |

nea-1390 | ZZ UKFY2, Fission Yields of Th, U, Np, Pu, Am, Cm, Cf Isotopes |

dlc-0164 | ZZ UNGER, Effective Dose Equivalent Data for Selected Isotopes |

dlc-0211 | ZZ UTXS6, MCNP Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1365 K. |

dlc-0256 | ZZ VIP-MAN, Computational Phantom |

nea-1264 | ZZ VITAMIN J/COVA, Covariance Matrix Data Library for Uncertainty Analysis |

dlc-0184 | ZZ VITAMIN-B6, Fine-Group Cross-Section Library from ENDF/B-VI.3 for Radiation Transport |

dlc-0041 | ZZ VITAMIN-C, 171 Neutron-Group, 36 Gamma-Group Coupled Cross-Section for Fusion, LMFBR Calculations |

dlc-0113 | ZZ VITAMIN-E, 174-Group Neutron, 38-Group Gamma Cross-Section in AMPX Format |

nea-1168 | ZZ VITAMIN-J/KERMA, Gas Production Cross-Sections, Neutron and Gamma Kerma in FOURACES Format |

dlc-0245 | ZZ VITAMINB7/BUGLEB7, Broad-Grp, Fine-Grp, Coupled N/Gamma Cross-Sec Lib derived from ENDF/B-VII.0 Nuclear Data |

nea-1870 | ZZ VITENDF70.BOLIB, ENDF/B-VII.0 Multi-Grp Coupled (199n +42gamma)X-Sec.Lib.in AMPX Fmt for Nuclear Fission Applications |

nea-1702 | ZZ VITENEA-E, AMPX 174-N,38-gamma multigroup X-sec.library for multidimensional radiation transport and dose evaluation |

nea-1703 | ZZ VITENEA-J, AMPX 175-N,42-gamma multigroup X-sect. library for nuclear fusion applications |

nea-1699 | ZZ VITJEF22.BOLIB, JEF-2.2 Multigr Coupled (199n + 42gamma) X-Section Lib. in AMPX Fmt for Nuclear Fission Applications |

nea-1801 | ZZ VITJEFF31.BOLIB,JEFF-3.1 Multigr Coupled (199n + 42gamma) X-Section Lib. in AMPX Fmt for Nuclear Fission Applications |

nea-1869 | ZZ VITJEFF311.BOLIB, JEFF-3.1.1 Multi-Group Coupled (199n + 42gamma) X-Sec Lib in AMPX Fmt for Nuclear Fission Applic. |

nea-1518 | ZZ WIMKAL-88, 69-Group KAERI WIMS Library for Thermal Reactors |

iaea1397 | ZZ WIMS-D/4LIB, 61, 64, 69, 76 and 79 energy groups WIMS-D/4 libraries |

nea-1207 | ZZ WIMS-LIB/JEF87, 69+1 Group WIMS-D Library from JEF-1 |

nea-0329 | ZZ WIMS-TRIGA, ZZ-WIMSLIB/IJS, WIMS Data Libraries |

dlc-0026 | ZZ WM-NRSM, Neutron and Gamma Group Cross-Section Library for Nuclear Rocket Shielding Calculations |

dlc-0174 | ZZ XCOM, Photon Cross-Section Library for Personal Computer |

iaea1257 | ZZ XG, Radionuclide Decay Parameters for Gamma and X-Ray Detector Calibration |

iaea1364 | ZZ-FENDL-2, Evaluated Nuclear Data Library for Fusion Neutronics Applications |

nea-1891 | ZZ-VITJEFF32.BOLIB, JEFF-3.2 Multi-Grp Coupled (199n+ 42gamma) X-Sec. Lib. in AMPX Fmt for Nuclear Fission Applications |