OECD Nuclear Energy Agency / L'Agence pour l'énergie nucléaire OECD-OCDE







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-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 ACTIVATE2007, 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
  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 ALICE2008, 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
  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-50, Plot of Scalar Flux and Dose Rates from ANISN Calculation
  ccc-0254 ANISN, 1-D Neutron Transport and Gamma Transport in Slab, Cylindrical, Spherical Geometry with Anisotropic Scattering
  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-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-1006 ASDIC, Fast Reactor Hexagonal 3 Component Fuel Pin Diffusion Coefficient
  nea-0661 ASNBILD, Generator of JCL and Data for Program ANISN on CDC Computer
  ccc-0126 ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport
  ests0803 ASR4, 4-Gage Anelastic Strain Recovery Analysis
  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
  nesc0417 ATHENA-4, Inelastic Scattering Form Factors for Woods-Saxon or Harmonic Potential
  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
  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
  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
  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
  csni2015 CABRI-WATER-LOOP, High burn-up fuel behaviour in RIA conditions
  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-2005, Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations
  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
  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 CEM03.01, 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 CFDLIB05, 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
  iaea1405 CHAINFINDER 2.16, search for transmutation chains under neutron irradiation
  ccc-0604 CHAINS-PC, Decay Chain Atomic Densities
  iaea1404 CHAINSOLVER 2.20, transmutation simulation of samples during irradiation in nuclear reactors
  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
  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
  ests0802 CHMTRNS, Non-Equilibrium Chemical Transport Code
  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
  nesc0313 CINDER, Depletion and Decay Chain Calculation for Fission Products in Thermal Reactors
  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, One, Two- and Three-Dimensional Coupled Neutral and Charged Particle Sn 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-0724 COG10, 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
  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 COMPLOT2007, 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
  uscd1203 COMTRANS, ENSDF Comment Font Case Translator
  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 CONVERT2007, 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-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
  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-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
  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
  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
  ccc-0257 DELFIC-TES, Gamma Doses from Nuclear Explosion Radioactive Clouds
  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
  iaea1308 DICTIN2007, 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
  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
  iaea0952 DIGA/NSL, 3 Regions Lattice Cell Neutron Flux Diffusion Calculation
  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-0533 DISKTRAN, Detector Response to Scalar Flux from DOT-4 Calculation
  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
  nesc9618 DISPPAK SUBPAK, MS FORTRAN Extended Subroutine Library
  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-1564 EASY-2005.1, European Neutron Activation System
  nea-1813 EASYQAD version 1.0, Visualization for Gamma and Neutron Shielding Calculations
  ests0288 EBQ, Steady-State Space Charge Transport in Cylindrical Geometry
  nea-0850 ECIS-06, 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
  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
  uscd1204 ENSDAT, Evaluated Nuclear Structure Drawings and Tables
  uscd1149 ENSDF ADDGAM, Adds Gammas to Adopted Data Sets
  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
  ests0651 ESPSD, Nuclear Power Plant Siting Database
  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 EVALPLOT2007, ENDF Plots Cross Section, Angular Distribution and Energy Distribution
  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
  ccc-0440 EXTREME, 2-D Discrete Ordinate System with Exponential Space Expansion
  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
  nea-1038 FAIR-DDX, Double Diffusion Cross-Sections Scattering Matrix Generated from ENDF/B-4 or JENDL-2
  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 Test L-14 on fuel coolant interaction and quenching
  nesc1095 FASTGRASS, Gaseous Fission Products Release in UO2 Fuel
  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
  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
  iaea0903 FEONAN, Flux Smoothing of Spectrometer System
  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 FIXUP2007, 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-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 FLUKA2008.3c, 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
  ests0016 FORECAST-V3.0, Forecast Regulatory Effects Cost Analysis
  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-2K6, 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
  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
  nea-1846 FSKY4C, Gamma Ray Skyshine Analysis Code
  nesc0666 FTA, Fault Tree Analysis for Minimal Cut Sets, Graphics for CALCOMP
  nesc9860 FTRANS, Radionuclide Flow in Groundwater and Fractured Rock
  nea-0068 FUELCYC-2-3, 2-D 2 Group U235 and U238 Fuel Depletion in Cylindrical Geometry
  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
  ccc-0628 GBANISN, ANISN Like 1-D Neutron and Gamma Transport with Group Band Fluxes
  iaea1362 GCASCAD, Gamma Production Cross Sections Statistical Model
  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-0601 GENII-1.485, GENII-LIN, Multipurpose Health Physics Code
  ccc-0648 GENII-S, Environmental Radiation Dosimetry System
  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, 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 GROUPIE2007, 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
  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)
  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
  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, Field Evaluation of Radiation Release from Nuclear Accident
  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-2009, 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-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-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-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-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-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-1765 IRPHE-HANDBOOK, International Handbook of Evaluated Reactor Physics Benchmark Experiments
  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-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-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
  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, a Java-based nuclear data display program
  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-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
  nesc0450 KENO, MultiGroup P1 Scattering Monte-Carlo Transport Calculation for Criticality, Keff, Flux in 3-D
  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-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 LEGEND2007, 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 LINEAR2007, 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
  nea-0860 LISA, Hazard Assessment of Nuclear Waste Disposal in Geological Formations
  uscd1214 LISTEF, ENDF/B Data File Summary List
  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
  ccc-0050 LRSPC, Proton High-Energy Loss in Matter
  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
  nea-0983 MARTHA, Nai(Tl) Gamma Scintillation Detector Response by Monte-Carlo
  csni1001 MARVIKEN-CFT, Marviken Full Scale Critical Flow Tests
  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-TREME, 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
  nea-1643 MCB1C, Monte-Carlo Continuous Energy Burnup Code
  csni2003 MCCI PROJECT, Molten Core Concrete Interaction Project
  nea-0452 MCDATA, MC**2 Cross-Sections Conversion for Programs CITATION, ANISN, DOT
  nea-1632 MCDSIM, Atmospheric Monte Carlo Dispersion Simulation
  ccc-0699 MCNP-DSP, Monte Carlo Neutron-Particle Transport Code with Digital Signal Processing
  ccc-0718 MCNP-POLIMI v1.0, Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities
  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-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 MERGER2007, 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
  ests0184 MINI-THESAURUS, Energy Data Base Subject Thesaurus Generator
  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 MIXER2007, 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-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
  ests1098 MPICH, Message Passing Interface (MPI) Subroutine Library for Parallel Computers and Networks
  ccc-0655 MRIPP, Magnetic Resonance Images (MRI) Data for Calibration of In Vivo Radionuclides Deposit Measurements
  nesc0798 MSF21/VTE21, Desalination Plant Heat, Mass Balance, Design, Cost Optimization
  nesc0508 MUCHA1, Fuel Rod Pair Thermohydraulics During LOCA and ECCSA 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-1176 MUTUAL, Nuclear Reactor Criticality Safety Analysis for Array System
  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
  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
  iaea1384 NKE 2.16, Nuclide Explorer tool for retrieving interactively detailed data on radionuclides properties
  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-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 2.3.13, Evaluation of Routine Radioactive Effluents from Nuclear Power Plants
  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
  psr-0284 NUFACE, Tokamak Fusion Reactor Nuclear Response Calculation
  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
  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-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
  ests0310 ORCON1, Steam Condenser Design with Circular Cross-Sections
  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
  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
  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
  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 PENELOPE2008.1, 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
  uscd1207 PHREEQC, Modeling of Geochemical Reactions, Calculation of pH, REDOX Potential
  nesc9674 PHREEQE, Modelling of Geochemical Reaction, Calculation of P-H, Redox Potential
  uscd1183 PHRQPITZ, Geochemical Calculation in Brines
  ccc-0160 PICA, Photon-Induced Medium-Range Nuclear Cascade Analysis by Monte-Carlo
  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
  iaea1426 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-RESULTS, Results of the Experiments Performed in the PMK-2 Facility for VVER Safety Studies
  nesc0469 PMS-1, Geometry and Multiple Scattering Correlation to Experimental Polarization Data by Monte-Carlo
  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
  psr-0208 POLLA, Resonance Parameter R-Matrix to S-Matrix Conversion by Reich-Moore Method
  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
  nea-1352 PRASMA, Risk Analysis of Off-Site Protection from Reactor Accidents
  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
  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
  nea-1173 PREP, Input Preparation for Monte-Carlo Program SPOP
  nea-1485 PREP-45, Input Preparation for CITATION-2
  iaea1379 PREPRO2007, 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
  ccc-0574 PRISIM, Plant Risk Status Information Management System
  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
  ests0585 Portable Instrumented Communication Library
  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
  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-0279 RAFFLE-V, General Geometry Neutron Transport by Monte-Carlo
  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-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-0553 RASCAL 3.0.5, Radiological Doses from Accidental Release to Atmosphere
  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-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 RECENT2007, 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 and Capture 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 RELABEL2007, 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
  nea-0932 RESENDD, Resonance Cross-Sections Calculation from ENDF/B-4 and ENDF/B-5
  ccc-0552 RESRAD, Residual Radioactive Material Guideline Implementation
  ests1225 RESRAD, 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
  psr-0312 RFUNC, Diffusion Cross-Sections Data Analysis for Resonance Region
  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
  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
  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
  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-1688 SACALC2B, Calculates the average solid angle for source-detector geometries
  nea-1078 SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System
  ccc-0517 SADDE-MOD1, Beta Spectra Evaluation Input Generator for Program VARSKIN
  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 SAPHIRE6.64, System Analysis Programs for Hands-on Integrated Reliability
  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-0750 SCALE 6, 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
  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
  nea-1755 SCIP, Radioactive Surface Contamination Investigation Program
  psr-0210 SCOPE, Shipping Cask Optimization and Parametric Evaluation
  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-1C0, 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
  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
  nesc9867 SHAMT, Steady-State Forced Convection Heat Transfer
  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 SIGMA1-2007, 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 SIXPAK2007, 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
  iaea1288 SPC, Nuclear Data Fits by B Spline Optimization
  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
  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
  ccc-0460 SPOT1, Gamma-Ray Dose Rate from Cylindrical Source Volume
  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, 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
  ests1041 STR92, Energy Deposition in Accelerator Ring Components
  nea-0993 STRADE, Stratified Random Design for Reactor Safety Analysis
  nesc0539 STRAP-2, Stress Analysis of Structure with Static 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
  iaea1361 SUNF, Simplified UNF Code, Fast Neutron Calculation by Unified Hauser-Feshbach Theory
  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
  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
  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-1.0, 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 TART2005, 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
  iaea1252 TEST, Sort, Delete, List ANISN and DOT Cross-Sections Library Data
  iaea1272 THACT-RR, Analysis of Thermal Hydraulics Transients in Research Reactor Core
  nea-0774 THALES, Thermohydraulic LOCA Analysis of BWR and PWR
  nea-1098 THARC-S, Rod Bundle Thermohydraulic Transients of LMFBR for Single Phase Conditions
  nea-0634 THERLIB, Library Generated for THERMOS from FACEL Library
  nesc9940 THERMIT, 3-D Thermo-Hydraulics of BWR and PWR
  nesc0184 THERMOS BRT-1, 1-D Integral Transport for Neutron Spectra, Slab and Cylinder
  nea-0043 THERMOS, Space-Dependent Thermal Flux in 1-D Slab or 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-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
  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
  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
  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
  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
  nea-0486 TOTEM, Demand Assessment for Nuclear Power Plants and Conventional Power Plants
  nesc1098 TOUGH, Unsaturated Groundwater Transport and Heat Transport Simulation
  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-4.3.3 & 4.4, 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
  ests0370 TSORT, Nuclear Plant Automatic Training Task Assignment
  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
  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
  nesc1088 UMIBIO, U Mill Bioassay Dosimetry Model
  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, 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
  ests0752 VCODE, Ordinary Differential Equation Solver for Stiff and Non-Stiff Problems
  ccc-0262 VCS, Radiation Protection Factors in Vehicles by Monte-Carlo
  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
  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
  iaea0932 VIRGIN2007, 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
  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
  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
  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-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
  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
  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
  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
  iaea0949 ZZ BROND, Evaluated Neutron Data Library in ENDF-6 Format
  nea-1401 ZZ BUC-1/BENCHMARK, NEACRP Benchmark Specifications for Burnup Criticality Calculation
  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
  iaea1256 ZZ CENPL-DLS, Discrete Level Schemes and Gamma Branching Ratios 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 ENDL82, 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
  nesc0447 ZZ ETOG-1-DATA, Cross-Section Library for Programs MUFT3, MUFT5, GAM1, GAM2 Generated from ENDF/B
  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
  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