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Catalog of Programs in alphabetical order


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
nesc9588 3DGEOELE, 3-D Nonlinear Least Square Fit
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
csni1015 ACHILLES, Heat Transfer in PWR Core During LOCA Reflood Phase
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
nesc0465 ADLER, ENDF/B Adler-Adler Resonance Parameter to Point Cross-Sections with Doppler Broadening
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-0146 ALICE91, Particle Spectra from Compound Nucleus Decay
ccc-0558 ALKASYS, Rankine-Cycle Space Nuclear Power System
nesc9658 ALPHA/AMPU, Radionuclide Radioactivity from Alpha Spectrometer Measurements
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
nesc0486 ANCON, Space-Independent Reactor Kinetics with Linear or Nonlinear Thermal Feedback
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-2000, Isotope Inventories from Neutron Irradiation, for Fusion Applications
nea-1343 ANITA-4, Isotope Inventories from Neutron Irradiation, for Fusion Applications
nea-1657 ANL-BPB, Argonne National Laboratory Code Center Benchmark Problem Book
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
nesc0858 APACHE, 2-D Chemical Reactive Fluid Flow Dynamic for CW Chemical Lasers
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
nesc0580 ASTEM, Evaluation of Gibbs, Helmholtz and Saturation Line Function for Thermodynamics Calculation
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
nesc0767 BEACON/MOD3, 1-D and 2-D 2 Phase Flow and Heat Transfer in Containment, LWR LOCA
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
csni0076 BETHSY/6.9C, Loss of residual heat removal system during mid-loop operation
csni0062 BETHSY/9.1B, Cold Leg Break Test
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
csni2018 BIP PROJECT, OECD/NEA Behaviour of Iodine Project
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-1187 BOREAS, Nuclear and Thermohydraulic LWR Burnup Simulation
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
csni2005 BUBBLER CONDENSER, OECD/NEA Bubbler Condenser Project
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, Flux Conversion from Program DOT to Currents for Program BREESE
psr-0131 CARP-82, Currents for Program BREESE-2 from ZZ CAD or DOT-4 Flux
ccc-0024 CARSTEP, Particle Flux on Space Vehicle in Van Allen Zone
nesc0482 CASCADE, Intranuclear Gamma Cascade Calculation for Particle Emission Probability
nesc0742 CASIM, High Energy Cascades in Shields of Arbitrary Geometry Using Monte-Carlo Method
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
nesc0892 CCC, Heat Flow and Mass Flow in Liquid Saturated Porous Media
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
nesc0548 CEBUG, 3-D Transient Hydraulics for Na H2O Reaction by Finite Elements Method
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
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
nesc9774 CHEMTRN, Chemical Species Transport in Groundwater System
nesc0611 CHILES, Singularity Strength of Linear Elastic Bodies by Finite Elements Method
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
nesc0540 CLOTHO, Mass Flow Data Calculation for Program PACTOLUS
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
nesc0188 CMPXMAT, Transfer Function Calculation for Linearized Differential Equation
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
nesc0432 COBRA, Transient Thermohydraulics Fuel Elements Clusters, Subchannel Analysis Method
nesc9978 COBRA-3C/RERTR, Thermohydraulic Low Pressure Subchannel Transients Analysis
nea-1614 COBRA-EN, Thermal-Hydraulic Transient Analysis of Reactor Cores
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-0777 COG 11, 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
nesc9994 COIFES, Structure Graphics for Finite Elements Method Using Hidden Line Technique
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
nesc0704 COMCAN, Fault Tree Analysis, Minimal Cut Sets for Common Cause Failure
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-4MOD3, 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
csni1023 CORA-13, Experiment on severe fuel damage, core degradation and quench
csni1024 CORA-W2, Experiment on Severe Fuel Damage for a VVER-type PWR
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-0067 COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons
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
nesc9577 CPDES2, Coupled 2-D Partial Differential Equation Solution
nesc9576 CPDES3, Coupled 3-D Partial Differential Equation Solution
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
nesc9678 CRI, 4-Processor VAX-11/780 Simulation of CRAY Multitasking System
nea-1734 CRISSUE-S, Neutronics/Thermal-hydraulics Coupling in LWR Technology
iaea0873 CRITIC, In-Core Fuel Management for CANDU PWR
nea-1681 CRITICALITYACCIDENTS, A Review of Criticality Accidents, 2000 Revision, LA-13638 in PDF format
nesc9829 CROSSPLOT-3/CON-3D, 3-D and Stereoscopic Computer-Aided Design Graphics
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
nesc9636 CUBESIM, Hypercube and Denelcor Hep Parallel Computer Simulation
nea-0507 CURFIT SURFIT, 2-D Polynomial Least Square Fit to Experimental Data
nesc9533 CURVE LSFIT, Gamma Spectrometer Calibration by Interactive Fitting Method
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-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-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
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
iaea1241 DNTM/R2D, 2-D Transport in X-Y Geometry
csni0071 DOEL2/SGTR, Steam Generator Tube Rupture incident at the DOEL 2, Westinghouse 2 loop PWR
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
nesc9833 DOT-BPMD, Non Linear Heat Transfer in 2-D Plane or Axisymmetric Structures
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
nesc0753 DRVOCR, Minimization of Nonlinear Function, Variable Metric Method, Derivative Calculation
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
nesc9910 DYNA-2D, 2-D Hydrodynamic Finite Elements Method Program with Interactive Rezoning
nesc9909 DYNA3D, 3-D Finite Elements for Dynamic Response of Inelastic Solids
ests0138 DYNA3D2000*, Explicit 3-D Hydrodynamic FEM Program
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-1875 EACRP-D2O-LATTICES, Compilation of reactor physics measurements in HWRs lattices
nea-1564 EASY-2010, European Neutron Activation System
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
nesc0881 ELEFUNT, Testing of Elementary Function Subroutines
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)
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
nesc1143 EPIPE, Static and Dynamic Piping System Analysis
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 2.0, 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
nesc0601 ERF/ERFC, Calculation of Error Function, Complementary Error Function, Probability Integrals
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
csni1026 ERSEC, investigation of the reflooding phase of a Loss of Coolant Accident
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-0449 ESTOQ, 1st Collision Source Calculation for Program DOT in R-Z Geometry
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
iaea0898 F5TAB, ENDF/B-4 FILE 5 Data Conversion to Tabulated Form
nesc9578 FACET, Radiation View Factor with Shadowing
csni1020 FALCON-ISP1, ISP-2, fission product and aerosol transport in primary coolant system and in the containment
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
csni1019 FARO/L-14, Test L-14 on fuel coolant interaction and quenching
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
nesc0909 FC,LSEI,WNNLS, Least-Square Fitting Algorithms Using B Splines
iaea1245 FDMXPC, ENDF/B Processing, with Reich-Moore and Adler-Adler Resonance Parameter Calculation
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
nesc0577 FFEARS, Laplace Equation Boundary Value Problems with Dielectrics, X-Y-Z and Axisymmetric Geometry
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-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
ests0022 FIREDATA, Nuclear Power Plant Fire Event Data Base
nea-0472 FIREFLY, X-Ray Diffraction Intensities for Powder Patterns
nea-0897 FISP-6, Fission Products Inventory and Energy Release in Irradiated Fuel
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
csni0058 FIST/4DBA1, BWR/4-218 Simulated Double-Ended Large Break Test
csni0057 FIST/6IB1, BWR/6 System Responses to Intermediate Break in Recirculation Suction Line LINE
csni0054 FIST/6MSB1, BWR/6 Main Steamline Double-Ended Break Test
csni0056 FIST/6PMC1, BWR/6-128 Isolation Valve (MSIV) Closure without Power Scram
csni0055 FIST/6SB1, BWR/6 Simulated Recirculation Line Break
csni0053 FIST/6SB2C, BWR/6 Recirculation Suction Line Break Test
csni0059 FIST/T1QUV, Simulated Failure to Maintain Water Level in BWR/6-218
csni0060 FIST/T23C, Simulated Failure to Maintain Water Level in BWR/6-218
nea-0894 FITOCO, Fine Group to Coarse Group Neutron Flux Conversion for Spectra Unfolding
csni0001 FIX-II/2032, BWR Pump Trip Experiment 2032, Simulation Mass Flow and Power Transients
csni0049 FIX-II/3025, BWR FIX-II Pump Trip Experiment 3025, Immediate Split Size Break
csni0050 FIX-II/3061, BWR FIX-II Pump Trip Experiment 3061, Large Split Break
csni0051 FIX-II/5052, BWR FIX-II Pump Trip Experiment 5052, Guillotine Break Simulation
csni0052 FIX-II/6261, BWR FIX-II Pump Trip Experiment 626, Transient Dryout Tests
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 FLUKA2011.2b.6, 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
nesc0766 FRANTIC-NRC, Accident Sequence and Event Tree Analysis for System Availability and Operation
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
nesc0048 FUGUE, Steady-State Temperature and Pressure Analysis in Closed Channels
nesc0610 FUNPACK-2, Subroutine Library, Bessel Function, Elliptical Integrals, Minimax Approximation
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
csni1008 G2/716, Westinghouse G2 Loop Test Facility
csni1009 G2/718, Westinghouse G2 Loop Test Facility
csni1010 G2/736, Westinghouse G2 Loop Test Facility
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
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
nesc0547 GAMB-1T, Group Constant Library from P1 or B1 Approximation Neutron Spectra in ANISN Format, DOT Format
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
nesc0185 GAMTEC-2, MultiGroup Constant for Homogeneous or Heterogeneous Core
iaea0832 GAMX, Ge(Li) and Si(Li) Gamma Spectra and X-Ray Spectra Unfolding
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
nesc0605 GAUSS-5 GAUSS-7, Evaluation of Ge(Li) Detector Gamma Spectra
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 2013/2.2, 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
nesc0711 GEOCOST-BC, Geothermal Power Plant Electricity Generator Cost, Thermodynamics Calculation
nesc9834 GEOTHER, 2-D Heat Transport and 2-Phase Fluid Flow in Porous Rock
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
nesc0298 GGC-4, MultiGroup Neutron Spectra and Broad Group Cross-Sections Calculation, P1, B1, B2, B3 Approximation
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
nesc0682 GNATS, Nonlinear Stress Analysis of 2-D and Axisymmetric Static Structure by Finite Elements Method
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
nesc0045 GRACE GRACE-1, MultiGroup Multi-Region Gamma Attenuation Gamma Dose in Slab Geometry
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
nesc0760 GRFPAK, Graphics for Pipe Joint Heat Flow and Stress Analysis Program Cortes
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)
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
nesc0710 HAMOC, Pressure Transients in Reactor Vessel Piping System after Accidents
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
nesc0434 HEATMESH, Geometry Data Generator for Heat Transfer Calculation in Axisymmetric System
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
nesc0672 HONDO, Time-Dependent Elastic and Inelastic Stress Analysis Using Finite Element Method
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
nesc0467 HRG-3, Slowing-Down Neutron Spectra Using P1 and B1 Approximation with Average Cross-Sections Calculation
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
csni0000 I.T.D., CSNI Integral Test Facility Validation Matrix
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-1486 ICSBEP-2013, International Criticality Safety Benchmark Experiment Handbook
nea-1326 IFF, Full-Screen Input Menu Generator for FORTRAN Program
nea-1594 IFPE/AEAT-IMC, Onset Gas Release and Grain Face Venting Rates in Fuels
nea-1596 IFPE/AECL-BUNDLE, Fission Gas Release and Burnup Analysis, PHWR Fuel
nea-1799 IFPE/AEKI-EDB-E110, Experimental Database of E110 Claddings under Accident Conditions
nea-1788 IFPE/BN-MOX-M109/D3, Belgonucleaire Beznau-1 PWR irradiated MOX Fuel Rod M109/D3
nea-1863 IFPE/BN-MOX-M510/D10, Belgonucleaire Beznau-1 PWR irradiated MOX Fuel Rod M510/D10
nea-1560 IFPE/BR3-HBFRHCP, BR-3 High Burnup Fuel Rod Hot Cell Program
nea-1705 IFPE/CAGR-UOX-SWELL, Fuel swelling Data Obtained from the AGR/Halden Ramp Test Programme
nea-1858 IFPE/CANDU-FIO-130, CANDU experiment FIO-130 Fuel Behaviour under LOCA Conditions
nea-1783 IFPE/CANDU-FIO-131, CANDU experiment FIO-131 Fuel Behaviour under LOCA Conditions
nea-1777 IFPE/CANDU-IRDMR, In-Reactor Diameter Measuring RIG EXP-FIO-118 and EXP-FIO-119 Fuel Behaviour under LOCA Conditions
nea-1615 IFPE/CEA-DEFECT FUEL, Experiments Irradiated at CEA Grenoble
nea-1626 IFPE/CNEA-MOX-RAMP, CNEA Power Ramp Irradiations with (PHWR) MOX Fuels
nea-1595 IFPE/CONTACT REV.1, PWR Fuel Performance Tests Siloe Reactor
nea-1806 IFPE/DEFEX, Studsvik DEFEX BWR fuel secondary defect formation as a consequence of primary defects
nea-1807 IFPE/DEFEX-II DEMO, BWR fuel primary defect and conditions leading to secondary failure of the cladding by hydriding
nea-1597 IFPE/DEMO-RAMP-I & II, Pellet Clad Interaction Behaviour, Fast Power Ramping
nea-1645 IFPE/EFE-RO, Experimental Fuel Elements RO89 and RO51 in TRIGA 14 MW Reactor (INR-Pitesti)
nea-1841 IFPE/EXP-BDL-406, performance of natural UO2 fuel irradiated at low linear powers in NRU
nea-1774 IFPE/FMDP-MOX4-5, Weapons-Derived MOX Fuel DOE FMDP Test Irradiations Capsules 4 & 5, Advanced Test Reactor (ATR)
nea-1599 IFPE/FUMEX-I, Data from OECD Halden Reactor Project for FUMEX-1 (Fuel Modelling at Extended Burnup)
nea-1720 IFPE/FUMEX-II/CASE27, 7 idealised cases for functional dependence of FGR predictions
nea-1625 IFPE/GAIN, Gadolinia Doped UO2 Fuel Behaviour Experiment
nea-1736 IFPE/GBGI, Grain-Bubble Gas Interlinkage
nea-1697 IFPE/HATAC R1, Fission Gas Release at High Burn-up, Effect of a Power Cycling
nea-1510 IFPE/HBEP REV.1, Battelle's High Burn-Up Effects Programme for Fuel Performance
nea-1546 IFPE/IFA-429, Fission Gas Release, Thermal Behaviour U02 Fuel, Halden Reactor
nea-1488 IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden
nea-1729 IFPE/IFA-507-TF3-TF5, Database For Transient Temperature Experiment Ifa-507
nea-1629 IFPE/IFA-508 & 515, PCMI Behaviour of Thin Cladding Rods, JAERI and HRP
nea-1778 IFPE/IFA-514/565, LWR MOX Fuel Irradiation Tests - HBWR Irradiation with the Instrument Rig, IFA-514/565 (JAEA) 6 rods
nea-1860 IFPE/IFA-519.9, Three PWR rods irradiated to 90 MWd/kg UO2
nea-1549 IFPE/IFA-533, Fuel Thermal Behaviour at High Burnup, Halden Reactor
nea-1684 IFPE/IFA-534.14REV1, fission gas release as a function of burnup at high power (52-55 MWd/kg)
nea-1548 IFPE/IFA-535, Fission Gas Release, Power Ramps, High Burnup Fuel
nea-1547 IFPE/IFA-562, Pellet Surface Roughness Effect on Thermal Performances and PCMI
nea-1803 IFPE/IFA-585, In-Reactor Creep Behaviour of Zircaloy-2 and Zircaloy-4 under Variable Loading Conditions
nea-1773 IFPE/IFA-591, JAEA Power Ramp Tests of MOX Fuel Rods IFA-591
nea-1772 IFPE/IFA-597-MOX, Hollow and solid MOX rods experiments
nea-1685 IFPE/IFA-597.3, centre-line temperature, fission gas release and clad elongation at high burn-up (60-62 MWd/kg)
nea-1861 IFPE/IFA-629.1, The Re-irradiation of MIMAS-MOX Fuel in IFA-629.1
nea-1862 IFPE/IFA-650.1 & .2, LOCA testing at Halden, Two experiments, IFA-650 series
nea-1555 IFPE/INTER-RAMP, Fast Power Ramps Failures of Unpressurised Fuel Rods
nea-1532 IFPE/KOLA-3, WWER-440 Fuel Performance Data from KOLA-3 NPP, FGR
nea-1766 IFPE/KOLA-3-MIR-RAMP, KOLA-3 MIR test temperature during ramp, FGR and pressure at EOL, Bu up to 55 MWd/kgUO2
nea-1710 IFPE/MT4-MT6A-LOCA, Large-break LOCA in-reactor fuel bundle materials tests at NRU
nea-1758 IFPE/NFIR-1, Clad creepdown, power history effect on fission product distribution (6 PWR rods 40-64 MWd/kg in BR-3)
nea-1741 IFPE/NOVOVORONEZH, operation factor data of the Novovoronezh VVER-1000 fuel assembly 4108 rods
nea-1724 IFPE/NSRR-FK1-2-3, Behaviour of 3 BWR segments FK-1, FK-2 & FK-3 under RIA test conditions in NSRR
nea-1622 IFPE/OSIRIS R3, 4 PWR Rods Irradiated in the CEA Osiris Reactor
nea-1556 IFPE/OVER-RAMP, Pellet Clad Interaction Failure Analysis, Power Ramps
nea-1776 IFPE/PRIMO-BD8, Belgonucleaire and SCK-CEN PRIMO Ramped MOX Fuel Rod BD8
nea-1696 IFPE/REGATE L10.3, FGR and Fuel Swelling during power transient at medium burn-up (SILOE reactor)
nea-1634 IFPE/RISOE-1, Fission gas release from high-burnup water reactor fuel
nea-1502 IFPE/RISOE-II, Fuel Performance Data from Transient Fission Gas Release
nea-1493 IFPE/RISOE-III, Fuel Performance Data from 3rd Risoe Fission Gas Release
nea-1722 IFPE/ROPE-I, BWR, 4 rods, Ringhals, investigates clad creep-out from Studsvik (1986-1993)
nea-1723 IFPE/ROPE-II, PWR rod over pressure experiment from Studsvik
nea-1310 IFPE/SOFIT, WWER-440 Fuel Thermal Performance and Fission Gas Release
nea-1623 IFPE/SPC-RE-GINNA, Full Length and Segmented Fuel Rodlet Irradiation in PWR
nea-1809 IFPE/STEED-I, Stored Energy / Enthalpy Determination from Studsvik
nea-1557 IFPE/SUPER-RAMP, PCI Failure Threshold for PWR and BWR Fuels
nea-1648 IFPE/TRANS-RAMP, Fuel behaviour data from PWR/BWR TRANS-RAMP I, II, IV experiments
nea-1536 IFPE/TRIBULATION R1, Fuel Rod Behaviour at High Burnup
nea-1738 IFPE/US-PWR-16X16, Lead Test Assembly Extended Burnup Demonstration Program
nea-1677 IFPE/ZAPOROSHYE-V1K, Zaporoshye VVER1000 fuel behaviour data (4-8 cycles, Burnup about 50 MWd/kgUO2)
ests0169 ILUCG2, 2-D Partial Differential Equations Asymmetric Matrix Solver
ests0170 ILUCG3, 3-D Partial Differential Equations Linear Asymmetric Matrix Solver
nesc0715 IMPAC-2, Dynamic Impact Analysis for 1-D Nonlinear Spring Shipping Container Model
ests0005 IMPACTS-BRC2.1, General Radiological Impacts Analysis
nesc0779 IMPORTANCE, Minimal Cut Sets and System Availability from Fault Tree Analysis
nesc9473 IMPSOR, 3-D Boundary Problems Solution for Thermal Conductivity Calculation
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
nesc9649 INGRID, 3-D Mesh Generator for Program DYNA3D and NIKE3D and FACET and TOPAZ3D
ccc-0185 INREM-EXREM-3, Time-Dependent Organ Doses from Isotope Inhalation and Ingestion
nea-0554 INSUL, Calculation of Thermal Insulation of Various Materials Immersed in He
nesc0590 INTEG INSPEC, Accident Frequencies and Safety Analysis for Nuclear Power Plant
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
nesc9480 INTERP, Lexical Analysis for Problems Oriented Language Development
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-1340 INVENT-STUDSVIK, Fission Products Abundances in U235, U238, Pu239 Samples
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
nea-1715 IRPHE-JAPAN, Reactor Physics Experiments carried out in Japan
iaea1415 IRPHE-KNK-II-ARCHIVE, KNK-II fast reactor documents, power history and measured parameters
nea-1660 IRPHE-SNEAK, KFK SNEAK Fast Reactor Experiments, Primary Documentation
nea-1876 IRPHE-VENUS-RECYCLE, Plutonium Recycling Physics Project Critical Experiments
nea-1661 IRPHE-ZEBRA, AEEW Fast Reactor Experiments, Primary Documentation
nea-1687 IRPHE/B&W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments
nea-1662 IRPHE/JOYO MK-II, JOYO MK-II core management and characteristics database
nea-1765 IRPHE2014-HANDBOOK, International Handbook of Evaluated Reactor Physics Benchmark Experiments
nea-1726 IRPhE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents
nea-1728 IRPhE-HTR-ARCH-01, Archive of HTR Primary Documents
nea-1764 IRPhE-TAPIRO-ARCHIVE, Fast neutron source reactor primary documents, reactor physics experiments
nea-1739 IRPhE/AVR, High Temperature Reactor Experience, Archival Documentation
nea-1759 IRPhE/BERENICE, effective delayed neutron fraction measurements
nea-1713 IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility
nea-1714 IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility
ests0003 IRRAS, Integrated Reliability and Risk Analysis System for PC
iaea1328 ISABEL EVA PACE-2, Evaporation Model with Intranuclear Cascade Input
nesc1034 ISDMS, Inel Scientific Data Management System
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
nesc9656 ITMETH, Iterative Routines for Linear System
ests0219 ITOUGH2, Inverse Modeling for TOUGH2 Multiphase Flow Simulators
ccc-0467 ITS, TIGER System of Coupled Electron Photon Transport by Monte-Carlo
csni1027 IVO-Loop Seal Facility (Air/Water), Two-phase behaviour of a PWR cold leg loop seal during LOCA accidents
csni1018 IVO-THERMAL MIXING, study mixing of emergency cooling water with primary water during LOCA accident
csni1028 IVO/AIR-WATER-CCFL, Air/water countercurrent flow limitation experiments with full-scale fuel bundle structures
nesc9583 JAC, 2-D Finite Element Method Program for Quasi Static Mechanics Problems by Nonlinear Conjugate Gradient (CG) Method
iaea0940 JADSPE, Multi-Channel Gamma Spectra Unfolding Program
nesc1058 JAKEF, Gradient or Jacobian Function from Objective Function or Vector Function
nea-1760 JANIS 3.4, 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
nesc0490 JOSHUA-SYSTEM, Data Base Management System for Batch and Interactive Operation
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
csni0004 LEIBSTADT/STP-2001, BWR/6, Reactor Core Isolation Cooling System Test
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
nesc0460 LIFE-1, Stress Analysis Swelling and Performance of Cylindrical Fuel Elements in Fast Reactors
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
csni0034 LOBI/A1-04R, Loop for Blowdown Investigation, PWR Double-Ended Cold-Leg Break Test
csni0035 LOBI/A1-06, Loop for Blowdown Investigation, PWR Double-Ended 2A Cold-Leg Break
csni0036 LOBI/A1-66, Loop for Blowdown Investigation, PWR Double-Ended 2A Cold-Leg Break
csni0037 LOBI/A2-77, Loop for Blowdown Investigation, PWR MOD2 Small Leak Programme Experiment
csni0038 LOBI/A2-81, Loop for Blowdown Investigation. PWR MOD2 1% Cold-Leg Break
csni0003 LOBI/B-R1M, Loop for Blowdown Investigation, PWR Single-Ended Cold-Leg Break Experiment B
csni0074 LOBI/BT-00, Simulation of a Loss of Feedwater Transient (LOFW)
nea-0623 LOCA-MARK-2, Fuel Temperature and Clad Temperature in HWR Steam Generator LOCA
csni0017 LOFT/L2-3, Loss of Fluid Test, 2nd NRC L2 Large Break LOCA Experiment
csni0016 LOFT/L2-5, Loss of Fluid Test, 3rd NRC L2 Large Break LOCA Experiment
csni0022 LOFT/L3-5, Loss of Fluid Test, 5th NRC L3 Small Break LOCA Experiment
csni0018 LOFT/L3-6, Loss of Fluid Test, 6th NRC L3 Small Break LOCA Experiment
csni0021 LOFT/L3-7, Loss of Fluid Test, 7th NRC L3 Small Break LOCA Experiment
csni0020 LOFT/L6-7, Loss of Fluid Test, Anticipated Transients with Multiple Failures
csni0070 LOFT/L8-2, Severe Core Transient Experiment
csni0019 LOFT/L9-3, Loss of Fluid Test, Anticipated Transients with Multiple Failures
csni0010 LOFT/LP-02-6, Loss of Fluid Test, 1st OECD Large Break Experiment
csni0012 LOFT/LP-FP-1B, Loss of Fluid Test, Fission Product Release Experiment
csni0013 LOFT/LP-FP-2, Loss of Fluid Test, Fission Product Release from Fuel
csni0007 LOFT/LP-FW-1, Loss of Fluid Test, PWR Response to Loss-of-Feedwater Transient
csni0002 LOFT/LP-LB-1, Loss of Fluid Test, Large-Break LOCA Experiment
csni0008 LOFT/LP-SB-1, Loss of Fluid Test, Small Hot Leg Break LOCA, Early Pump
csni0009 LOFT/LP-SB-2, Loss of Fluid Test, Small Hot Leg Break LOCA, Delayed Pump
csni0011 LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressureinjection System (HPIS)
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
csni0080 MARVIKEN-ATT, Marviken Aerosol Transport Test experiments
csni1001 MARVIKEN-CFT, Marviken Full Scale Critical Flow Tests
csni0078 MARVIKEN-FSCB-I, Marviken Full Scale Containment Blowdown experiments Series I
csni0079 MARVIKEN-FSCB-II, Marviken Full Scale Containment Blowdown experiments Series II
csni1033 MARVIKEN-JIT, Marviken Full Scale Jet Impingement Tests experiments
csni2008 MASCA, In-vessel phenomena during severe accidents
csni2010 MASCA-2, In-vessel phenomena during severe accidents
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
nesc9933 MATHDOC, VAX VMS on-Line SLATEC-3.0 Documentation System
ests0279 MATHEW/ADPIC, Air Concentration and Ground Deposition from Point Sources
nesc9851 MATLOC, Transient Non Linear Deformation in Fractured Rock
nea-0380 MATRA, Void Simulation in Steam and H2O Mixture Channel in Accident
nea-0448 MATTEO, BWR Subchannel Steady-State and Transient Thermohydraulics
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
nesc9907 MAZE, Input Generator for Program DYNA2D and NIKE2D
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
csni2003 MCCI PROJECT, Molten Core Concrete Interaction Project
csni2017 MCCI-2 PROJECT, Melt Coolability and Concrete Interaction Phase 2 Project
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-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
nesc9853 MMT, 1-D Radionuclide Groundwater Transport
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
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-0816 MUENSTER, 2-D R-Z Geometry Thermohydraulics Calculation for Pebble-Bed Reactor
nea-0933 MULTI-KENO, Criticality Safety Analysis by Monte-Carlo
nea-1041 MULTIPLET, Large Event Trees for Risk Assessment Calculation
nesc9684 MULTITASKER, Multitasking Kernel for C and FORTRAN Under UNIX
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, 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 II, 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
nesc9489 NACHOS2, Incompressible Viscous Fluid Dynamic
nesc0717 NAHAMMER, Pressure Transients in Na LMFBR Piping System, Linear Fluid Hammer Theory
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.
nesc0780 NALAP, Thermohydraulics for Na Cooled LMFBR after Pipe Rupture and Accidents
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
nea-1874 NEACRP-H2O-LATTICES, Compilation of reactor physics measurements in LWRs lattices
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
csni1011 NEPTUN/5007, PWR LOCA Cooling Heat Transfer Tests for Loft, Boil-Off Experiments
csni1012 NEPTUN/5050, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test
csni1013 NEPTUN/5052, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test
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)
nesc9831 NETFLO, 3-D Steady-State Ground-Water Flow in Heterogeneous Medium
nea-0823 NEUPAC, Experimental Neutron Spectra Unfolding with Sensitivities
nesc9923 NIKE2D, Analysis of Static and Dynamic Response of 3-D Solids
nesc9725 NIKE3D, Static and Dynamic Response of 3-D Solids
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
psr-0355 NJOY-94, General ENDF/B Processing System for Reactor Design Problems
psr-0368 NJOY-97, General ENDF/B Processing System for Reactor Design Problems
nea-1025 NJOY-UTILITIES-EIR, Utility Program EPLOTR, CPLOTR, SEPR, COMBR, DECAYR for NJOY
psr-0171 NJOY91, General ENDF/B Processing System for Reactor Design Problems
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
nesc0695 NMMSS, NMMSS Utility, Data Base Maintenance and Update
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
nesc0146 NPRFCCP, Fuel Cycle Cost and Economics for Multi-Region Reactor
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
psr-0014 O5S, Calibration of Organic Scintillation Detector by Monte-Carlo
nesc1125 OCA-P, PWR Vessel Probabilistic Fracture Mechanics
nesc0753 OCOPTR, Minimization of Nonlinear Function, Variable Metric Method, Derivative Calculation
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
csni2014 OLHF, Sandia Lower Head Failure of the reactor pressure vessel OECD/NEA Project
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
nesc9906 ORION, Postprocessor for Finite Elements Program NIKE2D and DYNA2D
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
nesc0699 ORSIM, Nuclear Fuel, Fossil Fuel Hydroelectric Power Plant Cost and Economics
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
csni0014 OTIS/220100, Once-Through Integral Systems, 2-Phase Natural Circulation and Reflux
csni0015 OTIS/220402, Once-Through Integral Systems, Cold Leg Small Break LOCA
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
csni0061 PACTEL-ITE06, VVER-440 natural circulation stepwise coolant inventory reduction
nesc0540 PACTOLUS, Nuclear Power Plant Cost and Economics by Discounted Cash Flow Method
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
csni2004 PAKS PROJECT, the fuel behaviour in accident conditions on the basis of analyses of the PAKS-2 event
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
csni1014 PATRICIA/GV-6, Steady State Steam Generator Test Facility
nesc9797 PATTER, Pattern Recognition Data Analysis
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
csni1002 PDHT-HP, Post Dryout Heat Transfer Experiments, Upflow and Downflow Conditions
csni1003 PDHT-LP, Low Pressure Post Dryout Loop, Upflow Conditions
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 PENELOPE2011, A Code System for Monte-Carlo Simulation of Electron and Photon 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
csni1025 PHEBUS/B9+, Degradation of a PWR Type Core during a severe fuel damage
csni1021 PHEBUS/test-218, Behaviour of a Fuel Rod Bundle during a Large Break LOCA Transient with a two Peaks Temperature History
nesc0454 PHENIX, 2-D MultiGroup Diffusion Fast Reactor Burnup Calculation and Fuel Cycle Analysis
iaea1327 PHENOM, Nuclear Level Densities of Excited Nuclei
nea-1857 PHITS-2.64, 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
csni0048 PIPER-1/PO-SB-7, SBLOCA Simulation of Break in BWR-6 Plant at PIPER1 GE BWR Simulator
nea-0482 PIXSE, Scattering Moments Calculation from Scattering Law
iaea1172 PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method
csni2001 PKL-1, Experimental data on boron dilution and loss of residual heat removal in mid-loop operation (during shutdown)
csni0072 PKL/K9, Refill and Reflood Experiment in a Simulated PWR Primary System (PKL)
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
iaea0916 PLOT-3D/BARC, Interactive 3-D Colour Plotting
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
nesc9692 PLOTLIB, Graphics Library for FR80 and TMDS and RJET Systems
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-1789 PMK2-VVER440-REPORTS, Final reports on the PMK-2 projects for VVER Safety Studies
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
nesc9680 POSSOL, 2-D Poisson Equation Solver for Nonuniform Grid
iaea1249 POTAUS, Stopping Power and Particle Ranges in Various Material
nesc0340 POWERCO, Nuclear Power Plant Electricity Cost and Economics
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
iaea0817 PROB, Transport Equation in Slab Geometry and Collision Probability by Overrelaxation Method
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
csni2200 PSB-VVER, Computer code validation for transient analysis of VVER and RBMK reactors project
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
nea-1780 PWR-MOX/UOX-TRANS, OECD/NEA US-NRC PWR MOX/UO2 Core Transient Benchmark
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
nea-0939 RANDOM, Random Number Generator with Large Cycle Length
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
nesc0889 RAS, Fault Tree Analysis, Reliability, Minimal Cut Sets for Common Cause Failure
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
csni2300 RASPLAV, Refine accident management strategies during a reactor core meltdown
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
csni1022 REBEKA, Behaviour of a Fuel Bundle Simulator during a Specified Heatup and Flooding Period Results
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-0761 RECTC/RECTCF, 2nd Order Elliptical Partial Differential Equation, Arbitrary Boundary Conditions
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
nesc0733 RELAP3B/MOD110, Flow Temperature Pressure Steam Quality in LWR after LOCA and Accidents
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
nesc0465 RESEND, Infinitely Dilute Point Cross-Sections Calculation from ENDF/B Resonance Parameter
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
csni1029 REWET, PWR LOCA accidents experiments
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
iaea0969 RHEIN, Modular System for Reactor Design Calculation
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
iaea0929 RICANT, Neutron Transport in X-Y Geometry for Isotropic Scattering
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
nea-1132 RKFB, Space-Independent Reactor Kinetics with Temperature Feedback
nesc0831 RO-75, Reverse Osmosis Plant Design Optimization and Cost Optimization
nea-1449 ROLAIDS-CPM, 1-D Slowing-Down by Collision Problems Method
csni2021 ROSA-2, Rig-of-safety Assessment Project
csni0039 ROSA-III/912, BWR Rig of Safety Assessment for LOCA, 5% Split Break Test
csni0040 ROSA-III/916, BWR Rig of Safety Assessment for LOCA, 50% Split Break Test
csni0041 ROSA-III/922, BWR Rig of Safety Assessment for LOCA, 5% Split Break Test
csni0047 ROSA-III/923, BWR Rig of Safety Assessment for LOCA
csni0042 ROSA-III/926, BWR Rig of Safety Assessment for LOCA, 20% Double-Ended Break
csni0043 ROSA-III/952, BWR Rig of Safety Assessment for LOCA, Reference MSL Break Test
csni0044 ROSA-III/971, BWR Rig of Safety Assessment LOCA, Loss of Offsite Power Transient
csni0045 ROSA-III/984, BWR Rig of Safety Assessment for LOCA, 2.8% Split Break Test
csni0046 ROSA-IV/SB-CL-18, Large Scale Test Facility, 5% Cold-Leg Break Test
csni0073 ROSA-IV/SB-CL-27, Large Scale Test Facility, Gravity-Driven Safety Injection
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
csni1000 S.E.T., CSNI Separate Effects Test Facility Validation Matrix
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
nesc0900 SALE-2D, 2-D Fluid Flow, Navier Stokes Equation Using Lagrangian or Eulerian Method
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, 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-0405 SAPHIRE 7.27, 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-1694 SATIF/CYCLO-RADSAFE, Health Physics and Radiological Safety of Cyclotrons 10-250 MeV
iaea0917 SC2N3N, (n-2n) and (n-3n) Cross-Sections Systematics
ccc-0785 SCALE 6.1.2, Modular system for criticality, shielding, source term, fuel depletion/decay, inventories, reactor physics
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
nesc0802 SCHAFF, Single-Phase Flow, Heat Transfer in Porous Media, Geothermal Energy System
nea-0994 SCINFI, Quenching Function of Beta-Ray Liquid Scintillation Detectors
psr-0267 SCINFUL, Scintillation Neutron Detector Response by Monte-Carlo
csni2019 SCIP, OECD/NEA Studsvik Cladding Integrity Project
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
csni0027 SEMISCALE/S-06-3(LV), Thermal and Hydraulic Phenomena with LOCA in U-Tube PWR
csni0028 SEMISCALE/S-06-3(SV), Thermal and Hydraulic Phenomena with LOCA in U-Tube PWR
csni0023 SEMISCALE/S-IB-3, Semiscale MOD-2A, 21.7% Communication Cold Leg Break LOCA Experiment
csni0024 SEMISCALE/S-PL-3E-LV, Scaled Reference 4-Loop PWR Experiment, Loss of Offsite Power Simulation
csni0025 SEMISCALE/S-PL-3E-SV, Scaled Reference 4-Loop PWR Experiment, Loss of Offsite Power Simulation
csni0077 SEMISCALE/TEST1011, Large break LOCA blowdown starting from isothermal conditions of the primary coolant loop
csni0026 SEMISCALE/UT-1, Semiscale MOD-2A, 10% Communication Cold Leg Break LOCA Experiment
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
csni2020 SETH-2, OECD/NEA SESAR Thermal-hydraulics Project
csni2002 SETH/PANDA, Three-dimensional gas flow distributions relevant to in-reactor containments under accidents conditions
csni2000 SETH/PKL, Countermeasures for two types of PWR accidents
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
nesc0795 SHOCK, Nonlinear Dynamic Structure Analysis, Spring and Mass Model, Runge-Kutta-Gill Method
nea-0538 SHOSPA-MOD, Hot Spot Factors for Fuel and Clad, Hot Channel Factors
iaea0826 SHOVAV, Space-Dependent and Time-Dependent Neutron Diffusion with Temperature Feedback in Slab Geometry
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
nea-1552 SINBAD ACCELERATOR, Shielding Benchmark Experiments
nea-1553 SINBAD FUSION, Neutronics Benchmark Experiments
nea-1517 SINBAD REACTOR, Shielding Benchmark Experiments.
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
csni1017 SMD/12R305C, Steady state critical flow in nozzles, medium to high pressure conditions
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
nesc0189 SNC, Sn Constant Calculation for Program DSN and TDC
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
nesc0764 SOERP, Statistics and 2nd Order Error Propagation for Function of Random Variables
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
csni0075 SPES/SP-FW-02, Total Loss Feedwater with Emergency Feed Water Delayed in SPES facility
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)
nesc0749 STEEP-4, Fusion Reaction Rates for Maxwellian and Slowing-Down Plasma Ion Distribution
nea-0575 STESTA, Steady-State State-Variable Profiles of Thermohydraulic Piping System
csni2007 STEX-II, International Steam Explosion Experimental Data Base
nesc9852 STFLO, Steady-State H2O Flow in Porous Media
nesc0652 STFODE-COLODE, 1st Order Stiff Ordinary Differential Equation System by Collocation Method
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
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 5.0, 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
nesc9838 TAC0-3D, 3-D Linear or Nonlinear, Steady-State or Transient Heat Transfer
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
nea-1737 TALYS-1.2, computes nuclear reactions cross-sections, yields and spectra via a comprehensive set of nuclear models
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 TART2012, 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
nesc9908 TAURUS, Postprocessor of 3-D Finite Elements Plots
csni0005 TBL/311, Two Bundle Loop Facility, Small Break in Recirculation Line
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
csni2016 THAI, OECD/NEA Thermal-hydraulics, Hydrogen, Aerosols, Iodine (ThAI) Project
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 BRT-1, 1-D Integral Transport for Neutron Spectra, Slab and Cylinder
nesc0184 THERMOS-ANL, 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
csni1016 THETIS, Single Phase Cooling, Forced and Gravity Reflood, Level Swell Experiments
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
csni0029 TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA
csni0030 TLTA/6432, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA
ests0551 TMAP4, Tritium Migration Analysis Program Version 4
csni2012 TMI-VIP, Three Mile Island Reactor Pressure Vessel investigation OECD/NEA Project
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
nesc1019 TOP-DRAWER, Histograms, Scatterplots, Curve-Smoothing
nesc9801 TOPAZ, 2-D Plane or Axisymmetric Heat Conduction Analysis
nesc9599 TOPAZ-3D, 3-D Steady-State or Transient Heat Transfer by Finite Element Method
nesc9669 TOPAZ-SNLL, Transient 1-D Pipe Flow Analysis
nesc9801 TOPAZ2D, 2-D Finite Element Method Heat-Transfer and Electrostatic and Magnetostatic (E&M) Potential Field Program
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
iaea0833 TREEDE, Point Fluxes and Currents Based on Track Rotation Estimator by Monte-Carlo Method
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
iaea0804 TRIFIDO, Decay Constant and Prompt Neutron Calculation from Pulsed Neutron Experiment
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
nesc0814 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
ccc-0547 TWODANT-SYS, DANTSYS3.0, 1-D, 2-D, 3-D MultiGroup Discrete Ordinate Method Transport
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
nea-1682 U3-U5-PU9-CRITICALS, Critical Dimensions of Systems containing U235, Pu239, and U233
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
csni1007 UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA
csni1004 UPTF/TEST5A/RUN063, Steam/Water Flow Phenom.Blowdown PWR Cold Leg Break LOCA
csni1005 UPTF/TEST8A/RUN112, Flow Patterns in Hot or Cold Leg, PWR Large Break LOCA
csni1006 UPTF/TEST8B/RUN111, Flow Patterns in Hot or Cold Leg, PWR Large Break LOCA
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-0522 VARSKIN 3 V3.1.0, Dose Calculation for Skin Contamination, with Sadde Input Generator
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
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
iaea0821 WELWING, Material Buckling for HWR with Annular Fuel Elements
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-2013, 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
nesc0765 ZONE, Finite Elements Method Quadrilateral and Triangular Mesh Generator for 2-D Axisymmetric Geometry
iaea1371 ZOTT99, Data Evaluation Using Partitioned Least-Squares
nea-0331 ZUBOK-2-3, Stability Region of Nonlinear 1st Order Differential Equation System by Lie-Series
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-1398 ZZ 3DLWRCT, 3-D LWR Rod Ejection and Rod Withdrawal Benchmarks
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-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
nea-1731 ZZ BFBT, OECD/NEA-US/NRC NUPEC BWR Full-size Fine-mesh Bundle Tests Benchmark
nea-1429 ZZ BIB-PU-RECY, Pu Recycling Bibliography
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-1401 ZZ BUC-1/BENCHMARK, NEACRP Benchmark Specifications for Burnup Criticality Calculation
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
nea-1551 ZZ BWRSB-FORSMARKS, Stability Benchmark Data from BWR FORSMARKS 1 and 2
nea-1454 ZZ BWRSB-RINGHALS1, Stability Benchmark Data from BWR RINGHALS-1
nea-1640 ZZ BWRTT, BWR Turbine Trip Transient Benchmark Based on Peach-Bottom 2
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
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-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 AVKER, Neutron Kerma Response Function Data Library
dlc-0011 ZZ DLC-11 RITTS, 121-Group Coupled Cross-Section for ANISN, DOT, MORSE
dlc-0012 ZZ DLC-12D POPLIB, Secondary Gamma Yields and Cross-Section Library for POPOP-4 Calculation
dlc-0013 ZZ DLC-13B, Resonance Cross-Section Group Constant Library for Tungsten and Depleted Pu
dlc-0014 ZZ DLC-14 AIR, Group Constant Library of Secondary Gamma Transport in Air for ANISN Calculation
dlc-0015 ZZ DLC-15 STORM-ISRAEL, Gamma Interaction Cross-Section Library in ENDF/B Format for Transport
dlc-0016 ZZ DLC-16 COBB, 123 Neutron-Group Cross-Section Library from ENDF/B for XSDRN Calculation
dlc-0017 ZZ DLC-17 NOX, 119-Group Coupled Cross-Section of Nitrogen, Oxygen, Air for MORSE
dlc-0018 ZZ DLC-18 NAB, 100 Neutron-Group Cross-Section Library of Na and Al for ANISN, DOT, MORSE Neutron Transport
dlc-0019 ZZ DLC-19 DECAYGAM, Isotope Gamma Energy Library for Spectrometry Evaluation
dlc-0021 ZZ DLC-21, X-Ray Attenuation Cross-Section Library from 0.1 KeV to 1 MeV
dlc-0023 ZZ DLC-23F CASK, 40-Group Neutron and Gamma Coupled Cross-Section for PWR Shipping Casks
dlc-0028 ZZ DLC-28, 73-Group Neutron and Gamma Coupled Cross-Section for CTR Transport Calculation
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-0006 ZZ DLC-6 GAMLIB, 99-Group Cross-Section Library by SUPERTOG Calculation from ENDF/B
dlc-0090 ZZ DOSCOV, 24-Group Covariance Data Library from ENDF/B-V for Dosimetry Calculation
nea-0827 ZZ DOSCROS, 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
nea-1606 ZZ ECN-BUBEBO, ECN-Petten Burnup Benchmark Book, Inventories, Afterheat
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
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
iaea1364 ZZ FENDL-2, Evaluated Nuclear Data Library for Fusion Neutronics Applications
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
nesc0844 ZZ FUELS-DATA, Data Library for LWR Fuel Behaviour for FRAP Program
nea-0878 ZZ GAMDAT-78, Gamma Decay Data of Radioisotopes
dlc-0071 ZZ GAMMON, Activation Data Library for Fusion Reaction
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-1255 ZZ GREAC-ECN-3 REAC-ECN-4, Neutron Reaction Cross-Sections Library for Fusion Reactors
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/F, 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
nea-1848 ZZ KALININ3, KALININ-3 Coolant Transient Benchmark
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
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
iaea1258 ZZ NDS-INDEX, Index to the IAEA-NDS Documentation Series
iaea1279 ZZ NMF-90, Database for Neutron Spectra Unfolding
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-1618 ZZ ORESUND, Nordic Mesoscale Dispersion Experiments over Land-Water-Land
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/38B, Group Constant Library from ENDF/B Fission Product Data for ORIGEN Calculation
nea-1881 ZZ OSKARSHAMN 2, Oskarshamn-2 (O2) BWR Stability Benchmark
iaea1423 ZZ PADF-2007, Proton Activation Data File in ENDF-6 format
nea-1746 ZZ PBMR-400, OECD/NEA PBMR Coupled Neutronics/Thermal Hydraulics Transient Benchmark - The PBMR-400 Core Design
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-0196 ZZ PR-EDB, Power Reactor Embrittlement Database
iaea1277 ZZ PRONDOS, Evaluations of Selected Neutron Activation Reactions for Dosimetry
nea-1849 ZZ PSBT, NUPEC PWR Sub-channel Bundle Tests Benchmark
dlc-0126 ZZ PVE, 38-Group P8 Photon Cross-Section Library for Gamma Radiation Transport
nea-1607 ZZ PWR-AXBUPRO-GKN, Measured Axial Burnup Profiles, NPP Neckarewstheim
uscd1219 ZZ PWR-AXBUPRO-SNL, Computed Axial Burnup Profile Database for PWR
nea-1554 ZZ PWR-MSLB, PWR Main Steam-Line Break Benchmarks, Coupled Neutronics Thermal-Hydraulics
dlc-0134 ZZ RADDECAY, Decay Data Library for Radiological Assessment
nesc9554 ZZ REAC-2, Nuclide Activation and Transmutation
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
nea-1617 ZZ SIESTA, Atmospheric Dispersion Experiment over Complex Terrain
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-0093 ZZ SKYPORT, Importance Function for Neutron and Gamma for Skyshine Dose from Accelerator
dlc-0188 ZZ SKYSDATA-KSU, Neutron and Gamma Skyshine Responses
dlc-0178 ZZ SNLRML, Dosimetry Cross-Section Recommendations
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-1428 ZZ U-BIB-RECY, U Recycling Bibliography
nea-1769 ZZ UAM-LWR, Uncertainty Analysis in Modelling, Coupled Multi-physics and Multi-scale LWR analysis
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.
nea-1693 ZZ V1000CT-1&2, VVER-1000 Main Coolant Pump Switching-on, Coolant Mixing Tests, Main Steam-Line Break Benchmarks
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/B, 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
nea-1610 ZZ WPNCS BENCHM REP, Published Articles and Reports on Criticality Safety
nea-1505 ZZ WPPR, Pu Recycling Benchmark Results
nea-1434 ZZ WPPR-FR-MOX/BNCMK, Benchmark on Pu Burner Fast Reactor
dlc-0174 ZZ XCOM, Photon Cross-Section Library for Personal Computer
iaea1257 ZZ XG, Radionuclide Decay Parameters for Gamma and X-Ray Detector Calibration