| nesc0374 |
1-DX, 1-D Diffusion for Fast Reactor MultiGroup Cross-Sections, Group Constant Collapsing |
| nesc0325 |
2-DB, 2-D MultiGroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search |
| nea-1250 |
2D-SEEP, 2-D Ground Water Flow in Permeable Geologic Media |
| nesc0806 |
2DEPEP, Partial Differencial Equation Solution and Eigenvalues for Potential and Diffusion Problems |
| nesc9739 |
2DFLOW, 2-D Drainage Winds and Diffusion Simulation |
| iaea1386 |
2GWIHLIB, Generation and Plot of Cross Sections for HYDMN |
| nesc0567 |
3-DB, 3-D MultiGroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup |
| nea-1250 |
3D-SEEP, 3-D Ground Water Flow in Permeable Geologic Media |
| nea-1732 |
3D-TRANS-2003, Workshop on Common Tools and Interfaces for Radiation Transport Codes |
| nesc9588 |
3DGEOELE, 3-D Nonlinear Least Square Fit |
| psr-0248 |
ABAREX, Optical Statistical Model Neutron Cross-Sections Using ABACUS and NEARREX |
| nea-0912 |
ABLEIT-TRANS, Isotope Concentration and Sensitivities on Cross-Sections Data |
| nea-1839 |
ACAB-2008, ACtivation ABacus Code |
| nea-0976 |
ACCULIB, Program Library of Mathematical Routines |
| ccc-0442 |
ACDOS3, Neutron Activation Activities and Dose Rates |
| nea-1072 |
ACFA, Isotope Activation of Coolant and Structure Materials |
| csni1015 |
ACHILLES, Heat Transfer in PWR Core During LOCA Reflood Phase |
| iaea0975 |
ACORNS, Covariance and Correlation Matrix Diagonalization |
| nea-0621 |
ACRO, Organ Doses from Acute or Chronic Radioactive Inhalation or Ingestion |
| ccc-0372 |
ACT-ARA, Time-Dependent Radiation Source Terms |
| nea-0511 |
ACTIV, Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment |
| iaea0960 |
ACTIV-JINR, Experimental Gamma Spectra Unfolding |
| iaea1380 |
ACTIVATE2010, Activation Cross Section by Combining Cross Section and Multiplier (ENDF Format) |
| ests0171 |
ADASAGE, ADA Application Development System |
| nea-0480 |
ADDELT, Scattering Law Correlation for Delta Function Phonon Spectra |
| nea-1708 |
ADEFTA 4.1, Atomic Densities for Transport Analysis |
| nea-1708 |
ADEFTA 4.1, Atomic Densities for Transport Analysis |
| psr-0190 |
ADENA, Fission Products Beta Spectra and Gamma Spectra in 19 Group from U235 Pu239 Mixture |
| nesc0465 |
ADLER, ENDF/B Adler-Adler Resonance Parameter to Point Cross-Sections with Doppler Broadening |
| nesc0908 |
AERIN, Organ and Tissue Doses from Radioactive Aerosols |
| ests0165 |
AES, Automated Construction Cost Estimation System |
| ccc-0360 |
AIRDIF, Neutron and Gamma Doses from Nuclear Explosion by 2-D Air Diffusion |
| nea-0001 |
AIREK-MOD, Time Dependent Reactor Kinetics with Feedback Differential Equation |
| nea-0002 |
AIREK-PUL, Periodic Kinetics Problems of Pulsed Reactors |
| iaea1274 |
AIREKMOD-RR, Reactivity Transients in Nuclear Research Reactors |
| nea-1130 |
AIRGAMMA, External Gamma-Ray Exposure from Radioactive Cloud |
| nesc0326 |
AIROS-2A, Space-Independent Reactor Kinetics and Space-Dependent Heat Transfer, Mass Transfer |
| ccc-0341 |
AIRSCAT, Dose Rate from Gamma Air Scattering by Single Scattering Approximation |
| ccc-0110 |
AIRTRANS, Time-Dependent, Energy Dependent 3-D Neutron Transport, Gamma Transport in Air by Monte-Carlo |
| nea-0590 |
AKIMA'S-SPLINE, Curve and Surface Fit of Uni-Variate and Bi-Variate Function |
| iaea1432 |
AL-SHIELDER, calculates shielding thickness of aluminum for any photon emitting radionuclide between 0.5 to 10 MeV |
| nea-0500 |
ALARM, Thermohydraulics of BWR with Jet Pumps During LOCA |
| nea-0705 |
ALARM-P1, PWR Thermohydraulics for ECCS During Blowdown |
| nea-1353 |
ALBEDO ALBEZ, Gamma and Neutron Attenuation in Air Ducts |
| nea-0108 |
ALCI, Homogeneous 2-D MultiGroup Neutron Diffusion in X-Z, R-Z, R-Theta Geometry with Criticality Search |
| ccc-0577 |
ALDOSE, Dose Rate from Alpha Disk Source in H20 |
| uscd1238 |
ALICE2011, Particle Spectra from HMS precompound Nucleus Decay |
| psr-0146 |
ALICE91, Particle Spectra from Compound Nucleus Decay |
| ccc-0558 |
ALKASYS, Rankine-Cycle Space Nuclear Power System |
| nesc9658 |
ALPHA/AMPU, Radionuclide Radioactivity from Alpha Spectrometer Measurements |
| ccc-0612 |
ALPHN, (Alpha, N) Neutron Production in High-Level Waste Canisters |
| nea-0585 |
ALPS, Solid-State Detector Alpha Spectra Unfolding |
| nesc0815 |
ALVIN, Diffusion and Integral Data Comparison and Sensitivity Analysis |
| nea-0675 |
AMALTHEE, Emission Spectra for N, D, H3, He3, He4 Induced Reactions |
| nea-0403 |
AMARA, Correlation of Nuclear Data to Integral Experiment by Lagrange Multipliers |
| nesc0562 |
AMDLIBAE, IBM 360 Subroutine Library, Special Function, Polynomials, Differential Equation |
| nesc0563 |
AMDLIBF, IBM 360 Subroutine Library, Eigenvalues, Eigenvectors, Matrix Inversion |
| nesc0564 |
AMDLIBGZ, IBM 360 Subroutine Library for Data Processing, Graphics, Sorting |
| iaea1251 |
AMICO, Cross-Sections Data for ANISN, DOT from WIMS Library |
| psr-0315 |
AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5 |
| uscd0795 |
AMRAW, Risk Assessment Method for Radioactive Waste Management |
| nesc0486 |
ANCON, Space-Independent Reactor Kinetics with Linear or Nonlinear Thermal Feedback |
| nea-1235 |
AND, Atomic Number Densities for Criticality Calculation |
| nea-0321 |
ANDROMEDA, 1-D Burnup for Fuel Cycle Analysis of FBR |
| nea-1798 |
ANGELO-LAMBDA, Covariance matrix interpolation and mathematical verification |
| nea-0633 |
ANIPLO-D50, Plot of Scalar Flux and Dose Rates from ANISN Calculation |
| ccc-0082 |
ANISN-E, 1-D Transport Program ANISN with Exponential Model |
| nea-0363 |
ANISN-FONTENAY, 1-D Planar, Spherical, Cylindrical Neutron Transport and Gamma Transport with Deep Penetration |
| ccc-0082 |
ANISN-JR, 1-D Transport Program ANISN with ZZ JSD Data and Flux Plot |
| ccc-0254 |
ANISN-ORNL, 1-D Neutron Transport & Gamma Transport in Slab, Cylindrical, Spherical Geometry with Anisotropic Scattering |
| ccc-0255 |
ANISN-W, 1-D Transport Calculation for Deep Penetration Problems |
| ccc-0514 |
ANISN/PC, MultiGroup 1-D Discrete Ordinates Transport with Anisotropic Scattering |
| nea-1638 |
ANITA-2000, Isotope Inventories from Neutron Irradiation, for Fusion Applications |
| nea-1343 |
ANITA-4, Isotope Inventories from Neutron Irradiation, for Fusion Applications |
| nea-1657 |
ANL-BPB, Argonne National Laboratory Code Center Benchmark Problem Book |
| nea-0470 |
ANSCLAD-1, Creep Strain in Fuel Pin Zircaloy Clad During Temperature Transient |
| nesc0529 |
ANVENT, Temperature Distribution and Pressure in Containment and Ice Condenser after LOCA for LWR |
| nesc9977 |
ANYOLS, Least Square Fit by Stepwise Regression |
| nesc0858 |
APACHE, 2-D Chemical Reactive Fluid Flow Dynamic for CW Chemical Lasers |
| nea-0546 |
APPLE, Plot of 1-D MultiGroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN |
| nea-0367 |
APPROX, 1-D and 2-D Function Approximation by Polynomials, Splines, Finite Elements Method |
| nea-0445 |
APS-2, Elastic Behaviour of Piping System |
| psr-0065 |
APSAI, Activation Calculation and Plot of Neutron Spectra, Gamma Spectra by ANISN |
| iaea1219 |
APUD-3.0, Off-Site Contamination Assessment from Accidental Release |
| ests1169 |
ARCON96, Radioactive Plume Concentration in Reactor Control Rooms |
| nea-0320 |
ARGO, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry from JAERI Fast-Set, ABBN, RCBN |
| nesc0152 |
ARGUS, Transient Temperature Distribution Cylindrical Geometry, Space-Dependent or Time-Dependent Heat Generator |
| nea-1368 |
ARIANNA-2, Sub-Compartment Thermo-Hydraulic Transients in LOCA |
| nea-0174 |
ARLEKIN, General Point Reactor Kinetics by Lie-Series Method |
| nesc0925 |
ARRRG/FOOD, Doses from Radioactive Release to Food Chain |
| nesc0738 |
ARSTEC, Nonlinear Optimization Program Using Random Search Method |
| nea-1581 |
ART MOD2, Fission Product Migration in Primary System and Containment |
| nea-0539 |
ASCOT-1, Thermohydraulics of Axisymmetric PWR Core with Homogeneous Flow During LOCA |
| nea-0661 |
ASNBILD, Generator of JCL and Data for Program ANISN on CDC Computer |
| ccc-0126 |
ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport |
| nesc0580 |
ASTEM, Evaluation of Gibbs, Helmholtz and Saturation Line Function for Thermodynamics Calculation |
| ccc-0417 |
AT123D, 1-D, 2-D, 3-D Transient Waste Transport Simulation in Groundwater |
| psr-0431 |
ATHENA_2D, Simulation Hypothetical Recriticality Accident in a Thermal Neutron Spectrum |
| ccc-0179 |
ATR, Radiation Transport Models in Atmosphere at Various Altitudes |
| iaea0906 |
AUJP, Optical Potential Parameter Search by CHI**2 Method |
| ccc-0519 |
AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors |
| psr-0008 |
AUTOJOM, Quadratic Equation Coefficient for Conic Volume, Parallelepipeds, Wedges, Pyramids |
| nea-1076 |
AVACOM-ETAP, Availability and Element Transient and Asymptotic Repair Process |
| nesc9700 |
AVPROG, Monte-Carlo Simulation of System Availability |
| nea-0861 |
AWE-1 AWE-2 BRUNA, Minimal Cut Sets of Logic Trees |
| nesc0191 |
AX-TNT, Super Prompt Critical Excursions in Spherical Geometry, Thermohydraulics |
| nea-0179 |
AXIFLUX, Cosine Function Fit of Experimental Axial Flux in Cylindrical Reactor |
| psr-0075 |
AXMIX, ANISN Cross-Sections Mixing, Transport Corrections, Data Library Management |
| psr-0297 |
AXMIX-PC, Cross-Sections Generator for ANISN, DOT from Different Sources |
| nesc9564 |
AYER, 2-D Thermal Conduction by Finite Element Method |
| nesc1020 |
BACFIRE, Minimal Cut Sets Common Cause Failure Fault Tree Analysis |
| uscd1158 |
BALANCE, Mass Transfer in Groundwater Aqueous Solution |
| nesc9677 |
BARMOM, Fission Barriers and Moments of Inertia |
| iaea0953 |
BASACF, Integral Neutron Spectra Adjustment and Dosimetry |
| nea-0636 |
BASKER, Isotropic Scattering Kernel Calculation Using VIWI |
| uscd1040 |
BAYESZ, S-Wave, P-Wave Resonance Level Spacing and Strength Functions |
| 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-2010, Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations |
| nea-1278 |
CALENDF-2010, 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 |
| psr-0131 |
CARP-82, Currents for Program BREESE-2 from ZZ CAD or DOT-4 Flux |
| ccc-0024 |
CARSTEP, Particle Flux on Space Vehicle in Van Allen Zone |
| nesc0482 |
CASCADE, Intranuclear Gamma Cascade Calculation for Particle Emission Probability |
| nesc0742 |
CASIM, High Energy Cascades in Shields of Arbitrary Geometry Using Monte-Carlo Method |
| psr-0262 |
CASKCODES, Program CAPSIZE Scope KWIKDOSE for Shipping Cask Shielding |
| nea-1195 |
CASKET, Thermal and Structural Analyses for Transport and Storage Cask |
| nea-0712 |
CASSANDRE, 2-D Reactor Dynamic FEM Program with Thermohydraulic Feedback |
| nea-1395 |
CASTHY, Statistical Model for Neutron Cross-Sections and Gamma-Ray Spectra |
| nesc0892 |
CCC, Heat Flow and Mass Flow in Liquid Saturated Porous Media |
| iaea1347 |
CCRMN, N, P, He4, D, H3, He3 Reaction Calculation for Medium-Heavy Targets |
| nesc9789 |
CDMS, Cost Data Management System Spread Sheet |
| iaea0920 |
CEBIS, 1-D 2 Group Diffusion Code for Reactor Calculation |
| nesc0548 |
CEBUG, 3-D Transient Hydraulics for Na H2O Reaction by Finite Elements Method |
| ests1071 |
CECP, Decommissioning Costs for PWR and BWR |
| nea-0553 |
CEDRAZAL, Steady-State Heat Transfer in HTR with Multifuel Region |
| psr-0532 |
CEM03.01, 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 |
CFDLIB, Computational Fluid Dynamics Library |
| nesc9537 |
CFEST-1.1, Coupled Fluid, Energy, Solute Transport in Ground-Water System |
| iaea1266 |
CFUP1, Neutron or Charged-Particle Reactions of Fissile Nuclei up to 33 MeV |
| 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 |
| nesc0638 |
CHECK-4, ENDF/B Utility, Structure Consistency Check and Format Check |
| uscd1208 |
CHECKR, ENDF/B Format Check |
| nea-1561 |
CHEMENGL/CHIMISTE, Chemical and Physical Properties of Elements |
| nea-1346 |
CHEMTARD, Simulation of Chemical Species Through Porous Media |
| nesc9774 |
CHEMTRN, Chemical Species Transport in Groundwater System |
| nesc0611 |
CHILES, Singularity Strength of Linear Elastic Bodies by Finite Elements Method |
| nea-0716 |
CHOLESK, Diffusion Calculation with 2-D Source in X-Y or R-Z Geometry |
| uscd1021 |
CHUCK-3, Nuclear Scattering Amplitude and Collision Cross-Sections by Coupled Channel |
| nea-0451 |
CICLON, Neutronics Calculation for PWR Transition Fuel Cycle Management |
| ccc-0755 |
CINDER 1.05, Actinide Transmutation Calculations Code |
| nesc0313 |
CINDER, Depletion and Decay Chain Calculation for Fission Products in Thermal Reactors |
| psr-0117 |
CINX, MINX Utility and SPHINX Utility, Library Data Collapsing |
| nesc9602 |
CIRCLE-SPLINE, 2-D, 3-D Spline Curve Fitting |
| nesc0387 |
CITATION, 3-D MultiGroup Diffusion with 1st Order Perturbation and Criticality Search |
| ccc-0643 |
CITATION-LDI2, 2-D MultiGroup Diffusion, Perturbation, Criticality Search, for PC |
| iaea1385 |
CITOPP,CITMOD,CITWI, Processing codes for CITATION Code |
| nea-0631 |
CLAPTRAP, Pressure Transients in LWR Containment During LOCA |
| nesc0540 |
CLOTHO, Mass Flow Data Calculation for Program PACTOLUS |
| iaea0883 |
CLUB, Cell Calculation PF Candu PWR Fuel Clusters |
| nea-0864 |
CLUHET, Steady-State Thermohydraulics of Rod Bundles with 1 Phase Flow |
| nea-0357 |
CLUP-77C, Collision Probability and Neutron Flux Distribution in Multi-Region Fuel Cluster |
| nea-0255 |
CLUS, Heat Transfer and Fuel Power in Liquid Cooled 7 Rod Fuel Elements Cluster |
| nesc0188 |
CMPXMAT, Transfer Function Calculation for Linearized Differential Equation |
| iaea1265 |
CMUP2, Reaction Cross-Sections for N, P, D, T, He3, He4 up to 50 MeV |
| ccc-0726 |
CNCSN 2009, One, Two- and Three-Dimensional Coupled Neutral and Charged Particle Sn Parallel Multi-Threaded Code System |
| nesc0873 |
COAST-4, Design and Cost of Tokamak Fusion Reactors |
| nesc0432 |
COBRA, Transient Thermohydraulics Fuel Elements Clusters, Subchannel Analysis Method |
| nesc9978 |
COBRA-3C/RERTR, Thermohydraulic Low Pressure Subchannel Transients Analysis |
| nea-1614 |
COBRA-EN, Thermal-Hydraulic Transient Analysis of Reactor Cores |
| ests0135 |
COBRA-SFS CYCLE3, Thermal Hydraulic Analysis of Spent Fuel Casks |
| nesc1091 |
COBRA-SFS, Thermal Hydraulics of Spent Fuel Storage System |
| nea-0294 |
CODAC, MultiGroup Cross-Sections Generation from ENDF/B for Monte-Carlo Program TIMOC |
| ccc-0777 |
COG 11, Multiparticle Monte Carlo Code System for Shielding and Criticality Use |
| psr-0375 |
COGAP, Nuclear Power Plant Containment Hydrogen Control System Evaluation Code |
| nea-0915 |
COGEND, Decay Data Generated in ENDF-6 Format |
| nesc9994 |
COIFES, Structure Graphics for Finite Elements Method Using Hidden Line Technique |
| nea-1126 |
COLLI-PTB, Neutron Fluence Spectra for 3-D Collimator System by Monte-Carlo |
| nea-0903 |
COLUMN, 1-D Migration for Various Physical Chemical Processes |
| psr-0286 |
COMBINE-PC, MultiGroup Neutron Cross-Sections in B1 or B3 Approximation from ENDF/B-5 |
| nesc0704 |
COMCAN, Fault Tree Analysis, Minimal Cut Sets for Common Cause Failure |
| nea-0340 |
COMET, Mechanical and Thermal Stress in Fuel Element Clad |
| psr-0343 |
COMIDA, Radionuclide Food Chain Model for Acute Fallout Deposition |
| nesc0482 |
COMNUC, Gamma Emission Neutron Emission Fission and Scattering Cross-Sections Using Hauser-Feshbach |
| psr-0302 |
COMNUC3B, Gamma Emission, Neutron Emission Fission and Scattering Cross-Sections Using Hauser-Feshbach |
| iaea0966 |
COMPAR, NJOY, GROUPIE, FLANGE-2, ETOG-3, XLACS MultiGroup Cross-Sections General Comparison |
| nesc0702 |
COMPARE, Transient Subcompartment Thermodynamics Analysis with 2 Phase Vent Flow |
| nesc0776 |
COMPARE-MOD1 COMPARE-MOD1A, 2 Phase Flow Thermodynamics, Pressure in LWR Containment |
| ests0023 |
COMPBRN3, Modelling of Nuclear Power Plant Compartment Fires |
| iaea1321 |
COMPLOT2010, Compare ENDF/B Plots of Reaction Data |
| nesc0649 |
COMQC, Quality Control Statistical Analysis for Means, Errors, Skewness, Kurtosis |
| nea-1578 |
COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System |
| nesc0663 |
COMRADEX-4, Doses from Radioactive Release, Meteorological Dispersion, Aerosol |
| iaea0928 |
COMTA, Ceramic Fuel Elements Stress Analysis |
| nesc0498 |
CONCEPT-5, Cost and Economics Analysis for Nuclear Fuel or Fossil Fuel Power Plant |
| ests0680 |
CONCHAS-SPRAY, Reactive Flows with Fuel Sprays |
| nea-0325 |
CONDENSE, Conversion of JAERI Fast-Set to ABBN Format with Self-Shielding |
| nea-0946 |
CONDN-63B, Thermohydraulics of Nuclear Power Plant Condenser |
| nea-0427 |
CONDOR-3, Local and Spectrum Dependent Burnup with Mesh-Wise Depletion |
| ccc-0416 |
CONDOS-II, Radiation Dose from Consumer Product Distribution Chain |
| nesc0433 |
CONTEMPT, LWR Containment Pressure and Temperature Distribution in LOCA |
| nesc0818 |
CONTEMPT-4MOD3, LWR Containment Long-Term Pressure Distribution and Temperature Distribution in LOCA |
| iaea1307 |
CONVERT2010, FORTRAN Program Converter for Different Computers |
| psr-0017 |
COOLC, Ne-213 Liquid Scintillation Detector Neutron Spectra Unfolding |
| nea-1305 |
COOLOD, Steady-State Thermal Hydraulics of Research Reactors |
| csni1023 |
CORA-13, Experiment on severe fuel damage, core degradation and quench |
| csni1024 |
CORA-W2, Experiment on Severe Fuel Damage for a VVER-type PWR |
| nea-0567 |
CORAN, PWR and BWR Containment Response to LOCA |
| iaea1226 |
CORD, PWR Core Design and Fuel Management |
| nesc0758 |
COREL, Ion Implantation in Solids, Range, Straggling Using Thomas-Fermi Cross-Sections |
| nesc0759 |
CORTES, Steady-State and Transient Heat Flow and Stress Analysis in Pipe Joints |
| nea-0383 |
COSANI-2, Gamma Doses from SABINE Calculation, Activity from ANISN Flux Calculation |
| nea-1375 |
COSIMA, BWR Core Performance Simulator |
| nea-0067 |
COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons |
| nea-0160 |
COSTANZA-AX, 1-D Neutronics and Thermodynamics of Liquid Cooled Reactor in Axial Geometry |
| nea-0160 |
COSTANZA-CYL, 1-D Neutronics and Thermodynamics of Liquid Cooled Reactor in Cylindrical Geometry |
| nea-0333 |
COSTANZA-RZ, 1-D Liquid Cooled Reactor Dynamic in R-Z Geometry |
| nea-0425 |
COSTANZA-XE, 2-D Pebble-Bed or Prismatic Fuel Elements HTR Dynamic in Cylindrical Geometry |
| nea-0398 |
COSTAX-BOIL, Transient Dynamic Analysis of BWR and PWR in Axial Geometry |
| nea-0533 |
COSTAX-BWR, Coupled Time-Dependent 2 Group Neutron Diffusion and 2 Phase Fuel Rod Coolant Flow |
| nea-0574 |
COVAL, Compound Probability Distribution for Function of Probability Distribution |
| nesc9577 |
CPDES2, Coupled 2-D Partial Differential Equation Solution |
| nesc9576 |
CPDES3, Coupled 3-D Partial Differential Equation Solution |
| ccc-0419 |
CRAC2, Reactor Accident Risk Assessment |
| nea-0463 |
CRACKLE, Fast Reactor Pu Fuel Management |
| nea-0057 |
CRAM-360, 1-D, 2-D Multi Group Diffusion with Keff Calculation or Criticality Search |
| nea-0718 |
CRAPONE, Optical Model Potential Fit of Neutron Scattering Data |
| nesc0638 |
CRECT, ENDF/B Utility, Data Correlation and Data Update |
| nea-0948 |
CRECT-J, Input Preparation of Evaluated Data in ENDF-4, ENDF-5 and ENDF-6 Formats |
| nesc9958 |
CREEP-80, Creep Analysis of Concrete Structure by Finite Element Method |
| nesc9678 |
CRI, 4-Processor VAX-11/780 Simulation of CRAY Multitasking System |
| nea-1734 |
CRISSUE-S, Neutronics/Thermal-hydraulics Coupling in LWR Technology |
| iaea0873 |
CRITIC, In-Core Fuel Management for CANDU PWR |
| nea-1681 |
CRITICALITYACCIDENTS, A Review of Criticality Accidents, 2000 Revision, LA-13638 in PDF format |
| nesc9829 |
CROSSPLOT-3/CON-3D, 3-D and Stereoscopic Computer-Aided Design Graphics |
| ccc-0518 |
CRRIS, Health Risk Assessment from Atmospheric Releases of Radionuclides |
| nea-1040 |
CRUNCH, Dispersion Model for Continuous Dense Vapour Release in Atmosphere |
| ccc-0233 |
CRYSTAL-BALL, Neutron Spectra Calculation from Activation Experiment with Error Estimate |
| nesc9636 |
CUBESIM, Hypercube and Denelcor Hep Parallel Computer Simulation |
| nea-0507 |
CURFIT SURFIT, 2-D Polynomial Least Square Fit to Experimental Data |
| nesc9533 |
CURVE LSFIT, Gamma Spectrometer Calibration by Interactive Fitting Method |
| nea-0247 |
CYGAS, 3-D Gamma Flux in Axial or Cylindrical Shields from Cylindrical Source |
| nea-0494 |
CYLDOS, Dose Rate in Cylindrical Shield from Cylindrical Source |
| nea-0371 |
CYLFUX, Fast Reactor Reactivity Transients Simulation in LWR by 2-D 2 Group Diffusion |
| nea-1416 |
D3D/D3E, 2-D, 3-D MultiGroup Neutron Diffusion in Rectangular, Cylindrical, Triangular-Z Geometry |
| ccc-0273 |
DACRIN, Dose in Respiratory Tract and Organs from Aerosol Inhalation |
| nesc0758 |
DAMG2, Ion Implantation in Solids, Energy Deposition Distribution with Recoils |
| nea-0151 |
DANCOFF-3, Dancoff Correction for Cylindrical Fuel Rod at H2O Gaps and for Fuel Clusters |
| nea-1516 |
DANCOFF-MC, Dancoff Correlation for Arbitrary Lattices by Monte-Carlo |
| nea-0103 |
DANG, Elastic and Direct Inelastic and Reaction Neutron Cross-Sections, Deformed Even-Even Nuclei |
| nea-0694 |
DANTE, Activation Analysis Neutron Spectra Unfolding by Covariance Matrix Method |
| ccc-0366 |
DASH, Void Tracing Sn and Monte-Carlo Coupling Program with Angular Fluxes from DOT Program |
| ests0357 |
DASH-FP, Multicomponent Time-Dependent Concentration Diffusion |
| nea-0646 |
DASQHE, Dancoff Correlation for Infinite Circular Rod Assembly in Square or Hexagonal Lattice |
| nesc9918 |
DASSL, Solution of Differential Algebraic Equation |
| nesc9493 |
DATING, Temperature for Spent Fuel Dry Storage |
| nea-0664 |
DCHAIN, Isotope Buildup and Isotope Decay by 1 Point Approximation |
| nea-1603 |
DCHAIN-SP 2001, Code System for Analyzing Decay and Build-up Characteristics of Spallation Products |
| ccc-0520 |
DCTDOS, Neutron and Gamma Penetration in Composite Duct System |
| nea-0229 |
DCXE, Time-Dependent Xe Diffusion in Non-Multiplying Slab |
| ests0848 |
DDASAC, Double-Precision Differential or Algebraic Sensitivity Analysis |
| iaea1290 |
DDCS, P, D, T, He3, He4 Reactions with 5 Particle Emission by Optical Model |
| nesc0640 |
DE/STE/INTRP, 1st Order Ordinary Differential Equation for Initial Value Problems |
| nea-0834 |
DEEBAR, Resonance Level Spacing Calculation by Dyson-Metha Optimum Statistics |
| ccc-0455 |
DEIS, Impact Measures of Low Level Radioactive Waste Disposal |
| nea-0446 |
DELIGHT-7, Point Reactivity Burnup for HTGR Lattice with P1 Neutron Scattering Approximation |
| nesc9681 |
DEM4-26, Least Square Fit for IBM PC by Deming Method |
| nesc0754 |
DEMONR, Monte-Carlo Shielding Calculation for Neutron Flux and Neutron Spectra, Teaching Program |
| ests0763 |
DENDRO, Cluster Analysis of Experimental Data with Tree Plot |
| nea-0840 |
DENZ, Dense Toxic or Explosive Gases Dispersion in Atmosphere |
| nea-0453 |
DEPCO-MULTI, Subcooled Decompression in PWR Primary System LOCA |
| psr-0523 |
DEPLETOR Version 2, provides depletion capability to the Purdue Advanced Reactor Core Simulator (PARCS) code |
| iaea0891 |
DIAG, 2-D Plotting Program for PDP-11/34 |
| nea-0672 |
DIAMANT-2, MultiGroup Neutron Transport with Anisotropic Scattering in Triangular Geometry |
| nesc0638 |
DICT-4, ENDF/B Utility, Section Table of Contents Generator |
| iaea1308 |
DICTIN2010, Reaction Index Generated for ENDF Format |
| ccc-0649 |
DIF3D 8.0/VARIANT8.0, 2-D 3-D Multigroup Diffusion/Transport Theory Nodal & Finite Difference Solver, Variational Method |
| ccc-0784 |
DIF3D10.0, Variational Nodal Methods, Finite Difference Methods to Solve N diffusion & Transport Theory Problems |
| iaea1269 |
DIFBAS, Spectra Unfolding of Ne213 P Recoil Detectors |
| nea-0667 |
DIFFAX, Axial Streaming for Hexagonal Lattices in Gas Cooled FBR, Slab Geometry Diffusion |
| nesc0737 |
DIFFUSER, 2-D and 3-D Diffuser Performance, Boundary Layer and Turbulent Flow |
| nea-0808 |
DIFFUSION-ACE, 3-D Neutron Diffusion by Leakage Iteration Method |
| nea-1067 |
DIFMOD, Radionuclide Leaching and Cement Corrosion in Brine |
| nesc9639 |
DIGLIB, General Graphics Subroutine Package for Different Computers |
| ests0243 |
DIGLIB, Multi Platform Graphics Subroutine Library |
| nea-0625 |
DINE, Neutron Flux, Neutron Dose Rate in Multi-Region Slab Reactor Shield by Removal Diffusion |
| nea-0298 |
DISCOUNT-G, Nuclear Power Program with Cost Analysis and Pu Production Optimization |
| nea-0643 |
DISCUS, Neutron Single to Double Scattering Ratio in Inelastic Scattering Experiment by Monte-Carlo |
| ccc-0170 |
DISDOS, Kerma in Model Man from External Gamma Source |
| ccc-0454 |
DISPERS, Radioactive Release into Surface Water and Ground Water |
| nesc0847 |
DISPL-1, 2nd Order Nonlinear Partial Differential Equation System Solution for Kinetics Diffusion Problems |
| nesc9532 |
DISPOSAL_SITE, Low-Level Radioactive Waste Storage Cost Analysis |
| nea-0184 |
DIXY-2, 2-D Homogeneous and Inhomogeneous Neutron Diffusion N X-Z, R-Z, R-Theta Geometry with Perturbation |
| nea-0391 |
DLS, 2-D Diffusion with Line-of-Sight Method for Cavities |
| iaea1241 |
DNTM/R2D, 2-D Transport in X-Y Geometry |
| csni0071 |
DOEL2/SGTR, Steam Generator Tube Rupture incident at the DOEL 2, Westinghouse 2 loop PWR |
| psr-0155 |
DOGS, Flux Plots of Radiation Transport Program Using DISSPLA |
| psr-0064 |
DOMINO, Coupling of Discrete Ordinate Program DOT with Monte-Carlo Program MORSE |
| iaea0961 |
DOMUS, Experimental 2-D Spectra Analysis |
| ccc-0650 |
DOORS, 1-,2-,3-dimensional discrete-ordinates system for deep-penetration neutron and photon transport |
| psr-0110 |
DOQDP ADOQ, Discrete Ordinate Quadrature Generator for Programs DOT and ANISN |
| nesc1146 |
DORIAN, Bayes Method Plant Age Risk Analysis |
| ccc-0543 |
DORT, 1-D 2-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration |
| ccc-0532 |
DORT-PC, 2-D Discrete Ordinates Transport System |
| nea-1711 |
DORTDAT2, Input-Making Support System for a Two-Dimensional SN Code, DORT |
| ccc-0624 |
DOSE-SGTR, Iodine Release During Steam Generator Tube Rupture (SGTR) in PWR |
| ccc-0536 |
DOSEFACTOR-DOE, Dose Rate Conversion Factors for Photon and Electron Exposure |
| iaea0922 |
DOSKMF2, Dose Rate Distribution in Co60 Gamma Irradiation Plants |
| ccc-0276 |
DOT-3.5, 2-D Neutron Transport, Gamma Transport Program DOT with New Space-Scaling |
| ccc-0320 |
DOT-4.2, 2-D Neutron Transport, Gamma Transport with Space Dependent Mesh and Quadrature |
| nesc9833 |
DOT-BPMD, Non Linear Heat Transfer in 2-D Plane or Axisymmetric Structures |
| ests0599 |
DPCT, Probabilistic Deterministic Contaminant Transport in Ground Water |
| nea-1506 |
DPOL3D, 2 Group, 3-D Core Transients and Steady State |
| uscd1234 |
DRAGON 3.05D, Reactor Cell Calculation System with Burnup |
| ccc-0647 |
DRAGON, Reactor Cell Calculation System with Burnup |
| uscd1237 |
DRAGON2PARTISN, Cross-Sections Data Generation for PARTISN4.0 |
| nea-1412 |
DRAWBS, NJOY Graphics Output of ENDF, PENDF, GENDF Data in GKS Format |
| iaea0885 |
DRUCK, Thermal, Mechanical Stress of PWR Fuel Rod During LOCA Blowdown |
| nea-0215 |
DRUCKSCHALE-44, Pressure and Temperature Transients in Blowdown Accident |
| nea-0839 |
DRUFAN-01/MOD2, Transient Thermohydraulics of PWR Primary System LOCA |
| nea-0457 |
DRUGEVO, Time-Dependent Containment Pressure and Temperature in BWR or PWR LOCA |
| nesc0753 |
DRVOCR, Minimization of Nonlinear Function, Variable Metric Method, Derivative Calculation |
| ests0637 |
DSEM, Radioactive Waste Disposal Site Economic Model |
| nesc0784 |
DSNP, Program and Data Library System for Dynamic Simulation of Nuclear Power Plant |
| psr-0251 |
DSNQUAD, Angular Quadrature Weights and Cosines for ANISN |
| nesc0209 |
DTF-4, 1-D MultiGroup Time-Independent Boltzmann Equation, Slab, Cylindrical, Spherical Geometry, Sn-Method, Pl-Method |
| nea-0269 |
DTF-G, Reactivity and Flux by 1-D Sn Method on Planar Cylindrical Spherical Geometry |
| nea-0322 |
DTF4-J, 1-D Neutron Transport with Anisotropic Scattering by Sn Method |
| nea-1671 |
DUCT-III, Design Code for Duct-Streaming Radiations |
| ccc-0453 |
DUST, Albedo Monte-Carlo Simulation of Neutron Streaming in Multilegged Square Concrete Ducts |
| ccc-0634 |
DUST-BNL, Radioactive Waste Transport from Container Leaks into Ground Water |
| nesc0579 |
DWARF, 1-D Few-Group Neutron Diffusion with Thermal Feedback for Burnup and Xe Oscillation |
| nea-1209 |
DWBA07/DWBB07, elastic scattering with nucleon-nucleon potential and DWBA for inelastic scattering |
| ccc-0383 |
DWNWND, Downwind Atmospheric Concentration and Dispersion by Gaussian Plume Model |
| nesc9872 |
DWUCK-4/5, Scattering Cross-Sections of Spin 0 and 1/2 and 1 Particles by DWBA |
| nea-1411 |
DYN3D/M2, Reactivity Transients in Light H2O Reactors with Hexagonal Geometry |
| nesc9910 |
DYNA-2D, 2-D Hydrodynamic Finite Elements Method Program with Interactive Rezoning |
| nesc9909 |
DYNA3D, 3-D Finite Elements for Dynamic Response of Inelastic Solids |
| ests0138 |
DYNA3D2000*, Explicit 3-D Hydrodynamic FEM Program |
| nesc0440 |
DYNAM, Once Through Boiling Flow with Steam Superheat, Laplace Transformation |
| nea-0090 |
DYNAMF, Time-Dependent Reactor Dynamics by Laplace Transformation |
| nea-0217 |
DYNAPS, Vibration Analysis of Piping System in Earthquake |
| ests1300 |
E3D, 3-D Elastic Seismic Wave Propagation Code |
| nea-1564 |
EASY-2010, 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 |
| ccc-0295 |
ELGATL, Calculation of Energy Spectra from Coupled Electron-Photon Slowing Down |
| nea-0435 |
ELIESE-3, Elastic, Inelastic, Reaction Cross-Sections, Polarization, by Hauser-Feshbach |
| iaea1223 |
ELPHIC-PC, Statistical Model Monte-Carlo Simulation of Heavy Ion Nuclear Reactions |
| ccc-0301 |
ELPHO, Muon, Electron, Positron Generator from Pions by Monte-Carlo with HETC Collision Data |
| nesc0546 |
EMERALD, Radiation Release and Dose after PWR Accident for Design Analysis and Operation Analysis |
| nesc0685 |
EMERALD-NORMAL, Routine Radiation Release and Dose for PWR Design Analysis and Operation Analysis |
| iaea1169 |
EMPIRE-II 2.18, Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections |
| uscd1235 |
ENDF-UTILITY-CODES, codes to check and standardize data in the Evaluated Nuclear Data File (ENDF) |
| iaea1402 |
ENDVER-ENDVER/GUI, The ENDF File Verification Support Package |
| uscd1149 |
ENSDF ADDGAM, Adds Gammas to Adopted Data Sets |
| uscd1149 |
ENSDF ALPHAD, Calculation Alpha Hindrance Factors |
| uscd1149 |
ENSDF AVETOOLS, Three statistical methods to calculate averages of experimental data with uncertainties |
| uscd1149 |
ENSDF BRICC, Interpolates Band-Raman internal conversion and electron-positron pair coefficients and E0 form factors |
| uscd1149 |
ENSDF DELTA, Gamma-Gamma Correlation Analysis |
| uscd1149 |
ENSDF ENSDAT, Graphics and Tables Generation from ENSDF Data |
| uscd1149 |
ENSDF FETCH, Indexing of ENSDF Files |
| uscd1149 |
ENSDF FMTCHK, Format Checking Program |
| uscd1149 |
ENSDF GABS, Absolute Gamma-Ray Intensities from ENSDF Data |
| uscd1149 |
ENSDF GTOL, Least Squres Fit of Gamma Spectra and Level Assignment |
| uscd1149 |
ENSDF HSICC, Interpolation Between Hager-Seltzer and Dragoun-Plajner-Schmutzler |
| uscd1149 |
ENSDF LOGFT, Beta-Decay log-ft and Partial Capture Calculation |
| uscd1149 |
ENSDF MEDLIST, Dose Rates from Nuclear Decay Data (X-ray intensities) |
| uscd1149 |
ENSDF NSDFLIB, Subroutine Library for ENSDF Programs |
| uscd1149 |
ENSDF PANDORA, Physics Checks on ENSDF Data |
| uscd1149 |
ENSDF PROCESSING CODES, Analysis and Utility Programs |
| uscd1149 |
ENSDF RADLST, Dose Rates from Nuclear Decay Data (decay of nuclei) |
| uscd1149 |
ENSDF RULER, Reduced Transition Problems Abilities Calculation |
| uscd1149 |
ENSDF SPINOZA, Tables of Levels, Decay, Gammy-Ray Data from ENSDF |
| uscd1149 |
ENSDF TREND, Tabulation of ENSDF Data |
| nea-0817 |
ENTOSAN, 640 Group Constant Calculation with Resonance from ENDF/B |
| nea-1686 |
ENTREE 1.4.0, BWR Core Simulation System for Space and Time Dependent Coupled Phenomena |
| iaea1285 |
EPICSHOW, Interactive Viewing of EPIC (Electron Photon Interaction Code) Data Library |
| nesc1143 |
EPIPE, Static and Dynamic Piping System Analysis |
| nesc0675 |
EPISODE, 1st Order Stiff or Non-Stiff Ordinary Differential Equation, Initial Value Problems |
| nesc0705 |
EPISODE-B, 1st Order Stiff or Non-Stiff Ordinary Differential Equation, Initial Value Problems |
| nesc0886 |
EQ-3 EQ6, Thermodynamics Equilibrium for Aqueous Solution Mineral System |
| iaea1202 |
EQUIVA, Few-Group Diffusion Parameter for PWR Reflector Region by 1-D Transport Calculation |
| nea-0261 |
EQUSTA, Thermodynamics Analysis and Mechanical Analysis for Fast Reactor Accident |
| nea-1683 |
ERANOS 2.0, Modular code and data system for fast reactor neutronics analyses |
| nea-0458 |
ERDBEBEN, Structure Displacements and Forces Under Earthquake Conditions |
| nea-0534 |
EREBUS, Burnup by 2-D MultiGroup Neutron Diffusion with Criticality Search |
| nesc0601 |
ERF/ERFC, Calculation of Error Function, Complementary Error Function, Probability Integrals |
| nea-0815 |
ERINNI, Emission Spectra for Multiple Cascades by Optical Model |
| nea-0515 |
EROS-2, Time-Dependent of Linear System by Inverse Laplace Transformation |
| nea-1676 |
ERRORJ, Multigroup covariance matrices generation from ENDF-6 format |
| csni1026 |
ERSEC, investigation of the reflooding phase of a Loss of Coolant Accident |
| nea-0341 |
ERUPT, 2-D 2 Group Fuel Management in R-Z Geometry with Fuel Shuffling |
| nea-0561 |
ESDORA, Continuous and Instantaneous Fission Products Release into Atmosphere |
| iaea1282 |
ESTAR PSTAR ASTAR, Stopping Power and Range of Electrons, Protons, Alpha |
| nea-0892 |
ESTIMA, Neutron Width Level Spacing, Neutron Strength Function of S- Wave, P-Wave Resonances |
| nea-0449 |
ESTOQ, 1st Collision Source Calculation for Program DOT in R-Z Geometry |
| nea-0984 |
ETHEL, Thermos Cross-Sections Library Generator Program |
| nea-0394 |
ETOA, ABBN MultiGroup Constants from ENDF/B for Fast Reactors |
| nea-1048 |
ETOBOX, Cross-Sections Library Generated from ENDF/B for Program BOXER |
| nesc0350 |
ETOE ETOE-2, Cross-Sections Library for Program MC**2 Generator from ENDF/B |
| nea-0630 |
ETOI, Format Conversion of Resonance Parameter from ENDF/B to Program IRESINT-3 Library |
| ccc-0107 |
ETRAN, Electron Transport and Gamma Transport with Secondary Radiation in Slab by Monte-Carlo |
| nea-0408 |
EURCYL, Mesh Generator for 3-D Intersections of Pressure Vessel Nozzles |
| nea-1094 |
EURDYN, Nonlinear Transient Analysis of Structure with Dynamic Loads |
| nea-0447 |
EUREKA, Reactivity Transients in LWR from Control Rod, Coolant Flow, Temperature |
| iaea1322 |
EVALPLOT2010, ENDF Plots Cross Section, Angular Distribution and Energy Distribution |
| nesc9952 |
EVENT, Explosive Transients in Flow Networks |
| nea-0893 |
EVGRP, Photo Production MultiGroup Cross-Sections Generated from ENDF/B-4 |
| psr-0465 |
EVNTRE, Code System for Event Progression Analysis for PRA |
| nea-0424 |
EXCURS, Heat Transfer Transients in Cylindrical Reactor Channel LOCA |
| nea-0228 |
EXCURS-3, Reactor Kinetics and Heat Transfer in Cylindrical Channel During Accident |
| iaea1273 |
EXCURS-3-RR, Kinetics of Research Reactor Reactivity Transient Analysis |
| iaea1211 |
EXIFON2.0, Neutron, Alpha, Proton, Gamma Emission Spectra |
| nesc0321 |
EXPALS, Least Square Fit of Linear Combination of Exponential Decay Function |
| nea-0311 |
EXPANDA-4, 1-D Neutron Diffusion in Slab, Spherical Geometry from JAERI Fast-Set with Criticality Calculation |
| nea-0312 |
EXPANDA-5, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry for 2 Region Fast Reactor Criticality |
| nea-0313 |
EXPANDA-6, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry from JAERI Fast-Set with PuO2 UO2 Mixture |
| nea-0315 |
EXPANDA-70, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry with Criticality Search |
| nesc0156 |
EXTERMINATOR-2, 2-D MultiGroup Neutron Diffusion in X-Y R-Z or R-Theta Geometry |
| psr-0237 |
EZVIDEO, DISSPLA Graphics Software Simulation on IBM PC |
| iaea0898 |
F5TAB, ENDF/B-4 FILE 5 Data Conversion to Tabulated Form |
| nesc9578 |
FACET, Radiation View Factor with Shadowing |
| csni1020 |
FALCON-ISP1, ISP-2, fission product and aerosol transport in primary coolant system and in the containment |
| ccc-0351 |
FALSTF, Neutron Flux and Gamma Flux Detector Response Outside Cylindrical Shields |
| nea-0592 |
FALT, Orientation of Double Coupled Earthquake Source with Given Amplitudes |
| ests0063 |
FAMREC, PWR Lateral Mechanical Fuel Rod Assembly Response |
| nea-0530 |
FANAC, Resonance Parameter by Multilevel Shape Analysis of Neutron Capture Yield Data |
| nea-0529 |
FANAL, Resonance Parameter by Multilevel Shape Analysis of Neutron Transmission Data |
| iaea0868 |
FAPCO, Evaluation of Flaws in Nuclear Power Plant Component Structures |
| nea-0617 |
FAPMAN-IC, LWR Fuel Cost Analysis with Program ORSIM Interface |
| nea-0693 |
FAPMAN-ORSIM, General Cost Optimization for System of Nuclear Power Plants |
| csni1019 |
FARO 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 |
| psr-0563 |
FEAST-METAL-V.1.0, Fuel Engineering and Structural analysis Tool |
| nesc1046 |
FED, Geometry Input Generator for Program TRUMP |
| iaea0830 |
FEDGROUP, Group Constant Library from ENDF/B, KEDAK, UKNDL |
| ests1121 |
FEHM, Finite Element Heat and Mass Transfer Code |
| nea-0930 |
FELPO, 2-D Minimization of Quadratic Functionals by Finite Elements Method |
| nea-0443 |
FEM-2D, 2-D MultiGroup Diffusion in X-Y Geometry |
| ests0198 |
FEM-3, Heavy Gas Dispersion Incompressible Flow |
| nea-0545 |
FEM-BABEL, 3-D MultiGroup Neutron Diffusion by Galerkin Method |
| nea-0566 |
FEM-RZ, 2-D MultiGroup Neutron Transport in R-Z Geometry, Eigenvalue and Fixed Source Problems |
| nea-1080 |
FEMAXI-6, Thermal and Mechanical Behaviour of LWR Fuel Rods |
| nea-0478 |
FEMB, 2-D Homogeneous Neutron Diffusion in X-Y Geometry with Keff Calculation, Dyadic Fission Matrix |
| ccc-0451 |
FEMWASTE FEMWATER, Finite Elements Method Waste Transport Through Porous Media |
| nesc1144 |
FEMWATER BLT, Water or Waste Transport in Soil |
| psr-0273 |
FERD-PC, Interactive Multichannel Neutron and Gamma Spectrum Matrix Unfolding |
| psr-0102 |
FERDO/FERD, Unfolding of Pulse-Height Spectrometer Spectra |
| psr-0145 |
FERRET, Least Square Fit to Nuclear Data and Reactor Physics Problems |
| ests0486 |
FESH, 2-D Multigroup Neutron Transport with Isotropic Scattering |
| ccc-0477 |
FEWA-FEMA, Finite Element Method Model of Materials Transport in Ground Water |
| nesc0577 |
FFEARS, Laplace Equation Boundary Value Problems with Dielectrics, X-Y-Z and Axisymmetric Geometry |
| nesc9844 |
FFSM, Long-Term Nuclear Waste Repository Site Simulation by Monte-Carlo |
| nea-1692 |
FFT-BM, Code Accuracy Evaluations with the 1D Fast Fourier Transform (FFT) Methodology |
| iaea1221 |
FIGA, Source Distribution Conversion from X-Y to R-Theta Geometry |
| iaea1181 |
FINEDAN, Dynamic Stress Analysis in 2-D X-Y and Axisymmetric Geometry |
| nea-0896 |
FINELM, MultiGroup Diffusion in 3-D by Finite Elements Method |
| nea-0310 |
FIP-DIG, 1-D Time-Dependent Fission Products Diffusion in Slab, Cylindrical, Spherical Geometry with Gaseous Precursor |
| nesc1092 |
FIRAC, Nuclear Power Plant Fire Accident Model |
| ests0022 |
FIREDATA, Nuclear Power Plant Fire Event Data Base |
| nea-0472 |
FIREFLY, X-Ray Diffraction Intensities for Powder Patterns |
| nea-0897 |
FISP-6, Fission Products Inventory and Energy Release in Irradiated Fuel |
| nea-0844 |
FISPET, MultiGroup Fission Spectra Calculation from ENDF/B |
| nea-0706 |
FISPIN, Isotope Buildup and Isotope Decay for Actinides, Fission Products, Structure Materials |
| nea-0182 |
FISPRO-2, Fast Neutron Capture Fission Product Cross-Sections by Hauser-Feshbach with Inelastic Scattering |
| csni0058 |
FIST/4DBA1, BWR/4-218 Simulated Double-Ended Large Break Test |
| csni0057 |
FIST/6IB1, BWR/6 System Responses to Intermediate Break in Recirculation Suction Line LINE |
| csni0054 |
FIST/6MSB1, BWR/6 Main Steamline Double-Ended Break Test |
| csni0056 |
FIST/6PMC1, BWR/6-128 Isolation Valve (MSIV) Closure without Power Scram |
| csni0055 |
FIST/6SB1, BWR/6 Simulated Recirculation Line Break |
| csni0053 |
FIST/6SB2C, BWR/6 Recirculation Suction Line Break Test |
| csni0059 |
FIST/T1QUV, Simulated Failure to Maintain Water Level in BWR/6-218 |
| csni0060 |
FIST/T23C, Simulated Failure to Maintain Water Level in BWR/6-218 |
| nea-0894 |
FITOCO, Fine Group to Coarse Group Neutron Flux Conversion for Spectra Unfolding |
| csni0001 |
FIX-II/2032, BWR Pump Trip Experiment 2032, Simulation Mass Flow and Power Transients |
| csni0049 |
FIX-II/3025, BWR FIX-II Pump Trip Experiment 3025, Immediate Split Size Break |
| csni0050 |
FIX-II/3061, BWR FIX-II Pump Trip Experiment 3061, Large Split Break |
| csni0051 |
FIX-II/5052, BWR FIX-II Pump Trip Experiment 5052, Guillotine Break Simulation |
| csni0052 |
FIX-II/6261, BWR FIX-II Pump Trip Experiment 626, Transient Dryout Tests |
| iaea1309 |
FIXUP2010, ENDF Format Redundant Cross-Sections Check |
| uscd1209 |
FIZCON, ENDF/B Cross-Sections Redundancy Check |
| nesc0395 |
FLAC FLAC-SI, Steady-State Flow and Pressure Distribution, 1-D Incompressible Flow Equation |
| nea-0636 |
FLAKER, Legendre Moments from Scattering Law Tables |
| nea-0551 |
FLANDES, Flange Design for He Circuits by Taylor-Forge Method |
| nesc0689 |
FLANGE-ORNL, Flanged Pipe Joint Stress Analysis, Internal Pressure, Moment Loads, Temperature |
| nesc0167 |
FLARE, 3-D BWR Reactivity and Power Distribution Appraisal Calculation |
| nea-0235 |
FLARE-JAERI, 3-D BWR and ATR Simulation |
| nea-0476 |
FLETU, Static Analysis of 3-D Pipeworks by Displacement Method |
| nesc9597 |
FLODIS, Thermal Response of FSV HTGR Core |
| nesc0246 |
FLOW-MODEL, Multichannel 2-D 2 Phase Flow for Open Matrix Flow BWR |
| nesc9592 |
FLOWPLOT2, 2-D, 3-D Fluid Dynamic Plots |
| nea-1833 |
FLUKA2011.2, Monte Carlo general purpose tool for calculations of particle transport and interactions with matter |
| psr-0196 |
FLYSPEC, Neutron Spectra Unfolding from Ne213 and Stilbene Scintillation Detectors |
| nea-0596 |
FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo |
| nesc0028 |
FOG-1-2-3, 1-D Few-Group Diffusion in Slab Cylindrical Spherical Geometry, Criticality Search, Buckling |
| nea-0669 |
FONTA, Radiation Release in Atmosphere and Deposition in Human Organs |
| nesc0174 |
FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients |
| psr-0092 |
FORIST, Ne-213 Scintillation Detector Neutron Spectra Unfolding |
| nea-0810 |
FORM-OTA, MultiGroup Constant for Epithermal Neutron Slowing-Down in Homogeneous Media |
| nesc0514 |
FORSIM, Solution of Ordinary or Partial Differential Equation with Initial Conditions |
| psr-0078 |
FORSIM-6, Automatic Solution of Coupled Differential Equation System |
| iaea1388 |
FOTELP-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 |
| nesc0694 |
FRAPCON2, Steady-State LWR Oxide Fuel Elements Behaviour, Fission Products Gas Release, Error Analysis |
| nesc0479 |
FREADM-1, Reactor Kinetics Thermohydraulics Calculation for Fast Reactor Accidents |
| nea-0692 |
FRELIB, Failure Reliability Index Calculation |
| nea-0982 |
FRETA-B, LWR Fuel Rod Bundle Behaviour During LOCA |
| nesc0301 |
FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements |
| nesc9659 |
FRTGEN, Fault Trees by Subtree Generator from Parent Tree for Program FTAP |
| nesc9659 |
FRTPLT, Fault Tree Structure and Logical Gates Plot for Program FTAP |
| nea-1846 |
FSKY4C, Gamma Ray Skyshine Analysis Code |
| nesc0666 |
FTA, Fault Tree Analysis for Minimal Cut Sets, Graphics for CALCOMP |
| nesc9659 |
FTAP, Minimal Cut Sets of Arbitrary Fault Trees |
| nesc9860 |
FTRANS, Radionuclide Flow in Groundwater and Fractured Rock |
| nea-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 |
| iaea1362 |
GCASCAD, Gamma Production Cross Sections Statistical Model |
| nea-1864 |
GEF, code for simulation of nuclear fission process |
| nesc0576 |
GEM, Fuel Cycle Cost and Economics for Thermal Reactor, Present Worth Analysis |
| nea-1652 |
GEM, Monte-Carlo Code for Simulating a Decaying Process of an Excited Nucleus |
| ests0742 |
GENAEA, Alpha Spectra Unfolding |
| nea-0606 |
GENDY, Reactor Dynamic Program with Variable Time Step Control |
| ccc-0737 |
GENII 2.10, Environmental Radiation Dosimetry System |
| ccc-0601 |
GENII-LIN, Multipurpose Health Physics Code |
| nea-0605 |
GENP-2, Program System for Integral Reactor Perturbation |
| nesc0711 |
GEOCOST-BC, Geothermal Power Plant Electricity Generator Cost, Thermodynamics Calculation |
| nesc9834 |
GEOTHER, 2-D Heat Transport and 2-Phase Fluid Flow in Porous Rock |
| uscd1210 |
GETMAT, ENDF/B Material Retrieval |
| nesc0887 |
GETOUT, Radioactive Release and Decay Chain Calculation for Nuclear Waste Disposal |
| nea-0584 |
GFX/GAMP1, Above-Ground Radiation Field from Terrestrial K, U, Th Gamma Emitters |
| nesc0298 |
GGC-4, MultiGroup Neutron Spectra and Broad Group Cross-Sections Calculation, P1, B1, B2, B3 Approximation |
| nea-0543 |
GGTC-ENEL, MultiGroup Neutron Spectra in P1, B1, B2, B3 Approximation and Thermos Calculation |
| nea-0073 |
GHT, 3-D Steady-State and Transient Heat Conduction |
| psr-0229 |
GIP, Group Organized Cross-Sections Library for ANISN, DOT |
| psr-0304 |
GIRAFFE, Isotope Release Analysis in LMFBR Fuel Elements Failure |
| psr-0192 |
GLUCS, Experimental Reaction Cross-Sections Evaluation for ENDF/B-5 |
| psr-0367 |
GMA, Generalized Least-Squares Cross-Sections Evaluation for ENDF Format |
| psr-0125 |
GNASH-FKK, FKK, Preequilibrium, Statistical Model Cross-Sections and Emission Spectra |
| nesc0682 |
GNATS, Nonlinear Stress Analysis of 2-D and Axisymmetric Static Structure by Finite Elements Method |
| iaea1271 |
GNOMER, Core Power Distribution by 1-D, 2-D, 3-D MultiGroup Neutron Diffusion |
| nea-0535 |
GOLIA-RK, Structure Stress for Isotropic Materials with Creep and Temperature Fields |
| nea-0550 |
GOMESH, Finite Elements Structure Plot with Triangular Mesh |
| nesc0045 |
GRACE GRACE-1, MultiGroup Multi-Region Gamma Attenuation Gamma Dose in Slab Geometry |
| nesc0046 |
GRACE-2, MultiGroup Multi-Region Gamma Attenuation Gamma Dose in Cylindrical or Spherical Geometry |
| uscd1211 |
GRALIB, DISSPLA Plot Routines Emulator |
| iaea1175 |
GRAP, Gamma-Ray Level-Scheme Assignment |
| nea-1043 |
GRAPE, System for Precompound and Compound Nuclear Reactions |
| nesc0624 |
GRAPH, Data Processing, Statistical Analysis, Correlations and Graphics |
| ests0075 |
GRASS-SST, Fission Products Gas Release and Fuel Swelling in Steady-State and Transients |
| nesc9911 |
GRAY CNVUFAC, Black-Body Radiation View Factors with Self-Shadowing |
| iaea0908 |
GRENADE, Green's Function Nodal Algorithm for Diffusion Equation |
| psr-0231 |
GRESS-3.0, FORTRAN Precompiler with Differentiation Enhancement |
| nea-0433 |
GRETEL, Ge(Li) Gamma Spectra Unfolding |
| nesc0760 |
GRFPAK, Graphics for Pipe Joint Heat Flow and Stress Analysis Program Cortes |
| ests0576 |
GRIDMAKER, 2-D, 3-D Finite Element Method Grid Generation for Ground Water and Pollutant Transport |
| nesc0620 |
GROUP-2, Atomic and Molecular Lattice Vibrations, Group Theory and Symmetry |
| iaea0849 |
GROUPIE2010, Bondarenko Self-Shielded Cross Sections from ENDF/B |
| nea-1111 |
GROUPXS, MF6 Format ENDFB-6 Continuum Region Diffusion Cross-Sections Processing |
| psr-0321 |
GRPANL, Ge Gamma and Alpha Detector Spectra Unfolding |
| ccc-0774 |
GRSAC, Graphite Reactor Severe Accident Code |
| nea-1690 |
GRTUNCL-3D/R-THETA-Z, Code to Calculate Semi Analytic First Collision Source and Uncollided Flux in an R-theta-Z grid |
| ccc-0721 |
GRTUNCL3D, Code to Calculate Semi Analytic First Collision Source and Uncollided Flux (X, Y, Z) |
| ccc-0276 |
GRUNCLE, 1st Collision Source Calculation for Program DOT |
| nesc9845 |
GSM, Columbia-Plateau Geologic Repository Site Long-Term Evolution Simulation |
| nea-1400 |
GTM-1, Radionuclide Transport Through Ground Water |
| nesc0618 |
GTR2 GAPCON-THERMAL2, Steady-State Fuel Rod Thermal Behaviour and Fission Products Gas Release |
| nea-1820 |
GTSP, automatic ultrasonic inspection of Guide Tube Support
Pin in nuclear power plants |
| ccc-0697 |
GUI2QAD, Graphical Interface for QAD-CGPIC, Point Kernel for Shielding Calculations |
| nea-0876 |
H2O, Calculation of Thermodynamics Properties of Steam and H2O |
| nea-0682 |
H2OTP, Temperature Dependent and Pressure Dependent Thermodynamics Properties, Transport Properties of H2O |
| nesc0443 |
HAA3B, Heterogeneous Aerosol Transport after LMFBR Accidents, Lognormal Size Distribution |
| nesc0797 |
HAARM, Time-Dependent Diffusion and Deposition of Radioactive Aerosols, LMFBR Accidents |
| 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 2.07.1, 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-2010, International Criticality Safety Benchmark Experiment Handbook |
| nea-1326 |
IFF, Full-Screen Input Menu Generator for FORTRAN Program |
| nea-1594 |
IFPE/AEAT-IMC, Onset Gas Release and Grain Face Venting Rates in Fuels |
| nea-1596 |
IFPE/AECL-BUNDLE, Fission Gas Release and Burnup Analysis, PHWR Fuel |
| nea-1799 |
IFPE/AEKI-EDB-E110, Experimental Database of E110 Claddings under Accident Conditions |
| nea-1788 |
IFPE/BN-MOX-M109/D3, Belgonucleaire Beznau-1 PWR irradiated MOX Fuel Rod M109/D3 |
| nea-1863 |
IFPE/BN-MOX-M510/D10, Belgonucleaire Beznau-1 PWR irradiated MOX Fuel Rod M510/D10 |
| nea-1560 |
IFPE/BR3-HBFRHCP, BR-3 High Burnup Fuel Rod Hot Cell Program |
| nea-1705 |
IFPE/CAGR-UOX-SWELL, Fuel swelling Data Obtained from the AGR/Halden Ramp Test Programme |
| nea-1858 |
IFPE/CANDU-FIO-130, CANDU experiment FIO-130 Fuel Behaviour under LOCA Conditions |
| nea-1783 |
IFPE/CANDU-FIO-131, CANDU experiment FIO-131 Fuel Behaviour under LOCA Conditions |
| nea-1777 |
IFPE/CANDU-IRDMR, In-Reactor Diameter Measuring RIG EXP-FIO-118 and EXP-FIO-119 Fuel Behaviour under LOCA Conditions |
| nea-1615 |
IFPE/CEA-DEFECT FUEL, Experiments Irradiated at CEA Grenoble |
| nea-1626 |
IFPE/CNEA-MOX-RAMP, CNEA Power Ramp Irradiations with (PHWR) MOX Fuels |
| nea-1595 |
IFPE/CONTACT REV.1, PWR Fuel Performance Tests Siloe Reactor |
| nea-1806 |
IFPE/DEFEX, Studsvik DEFEX BWR fuel secondary defect formation as a consequence of primary defects |
| nea-1807 |
IFPE/DEFEX-II DEMO, BWR fuel primary defect and conditions leading to secondary failure of the cladding by hydriding |
| nea-1597 |
IFPE/DEMO-RAMP-I & II, Pellet Clad Interaction Behaviour, Fast Power Ramping |
| nea-1645 |
IFPE/EFE-RO, Experimental Fuel Elements RO89 and RO51 in TRIGA 14 MW Reactor (INR-Pitesti) |
| nea-1841 |
IFPE/EXP-BDL-406, performance of natural UO2 fuel irradiated at low linear powers in NRU |
| nea-1774 |
IFPE/FMDP-MOX4-5, Weapons-Derived MOX Fuel DOE FMDP Test Irradiations Capsules 4 & 5, Advanced Test Reactor (ATR) |
| nea-1599 |
IFPE/FUMEX-I, Data from OECD Halden Reactor Project for FUMEX-1 (Fuel Modelling at Extended Burnup) |
| nea-1720 |
IFPE/FUMEX-II/CASE27, 7 idealised cases for functional dependence of FGR predictions |
| nea-1625 |
IFPE/GAIN, Gadolinia Doped UO2 Fuel Behaviour Experiment |
| nea-1736 |
IFPE/GBGI, Grain-Bubble Gas Interlinkage |
| nea-1697 |
IFPE/HATAC R1, Fission Gas Release at High Burn-up, Effect of a Power Cycling |
| nea-1510 |
IFPE/HBEP REV.1, Battelle's High Burn-Up Effects Programme for Fuel Performance |
| nea-1546 |
IFPE/IFA-429, Fission Gas Release, Thermal Behaviour U02 Fuel, Halden Reactor |
| nea-1488 |
IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden |
| nea-1729 |
IFPE/IFA-507-TF3-TF5, Database For Transient Temperature Experiment Ifa-507 |
| nea-1629 |
IFPE/IFA-508 & 515, PCMI Behaviour of Thin Cladding Rods, JAERI and HRP |
| nea-1778 |
IFPE/IFA-514/565, LWR MOX Fuel Irradiation Tests - HBWR Irradiation with the Instrument Rig, IFA-514/565 (JAEA) 6 rods |
| nea-1860 |
IFPE/IFA-519.9, Three PWR rods irradiated to 90 MWd/kg UO2 |
| nea-1549 |
IFPE/IFA-533, Fuel Thermal Behaviour at High Burnup, Halden Reactor |
| nea-1684 |
IFPE/IFA-534.14REV1, fission gas release as a function of burnup at high power (52-55 MWd/kg) |
| nea-1548 |
IFPE/IFA-535, Fission Gas Release, Power Ramps, High Burnup Fuel |
| nea-1547 |
IFPE/IFA-562, Pellet Surface Roughness Effect on Thermal Performances and PCMI |
| nea-1803 |
IFPE/IFA-585, In-Reactor Creep Behaviour of Zircaloy-2 and Zircaloy-4 under Variable Loading Conditions |
| nea-1773 |
IFPE/IFA-591, JAEA Power Ramp Tests of MOX Fuel Rods IFA-591 |
| nea-1772 |
IFPE/IFA-597-MOX, Hollow and solid MOX rods experiments |
| nea-1685 |
IFPE/IFA-597.3, centre-line temperature, fission gas release and clad elongation at high burn-up (60-62 MWd/kg) |
| nea-1861 |
IFPE/IFA-629.1, The Re-irradiation of MIMAS-MOX Fuel in IFA-629.1 |
| nea-1862 |
IFPE/IFA-650.1 & .2, LOCA testing at Halden, Two experiments, IFA-650 series |
| nea-1555 |
IFPE/INTER-RAMP, Fast Power Ramps Failures of Unpressurised Fuel Rods |
| nea-1532 |
IFPE/KOLA-3, WWER-440 Fuel Performance Data from KOLA-3 NPP, FGR |
| nea-1766 |
IFPE/KOLA-3-MIR-RAMP, KOLA-3 MIR test temperature during ramp, FGR and pressure at EOL, Bu up to 55 MWd/kgUO2 |
| nea-1710 |
IFPE/MT4-MT6A-LOCA, Large-break LOCA in-reactor fuel bundle materials tests at NRU |
| nea-1758 |
IFPE/NFIR-1, Clad creepdown, power history effect on fission product distribution (6 PWR rods 40-64 MWd/kg in BR-3) |
| nea-1741 |
IFPE/NOVOVORONEZH, operation factor data of the Novovoronezh VVER-1000 fuel assembly 4108 rods |
| nea-1724 |
IFPE/NSRR-FK1-2-3, Behaviour of 3 BWR segments FK-1, FK-2 & FK-3 under RIA test conditions in NSRR |
| nea-1622 |
IFPE/OSIRIS R3, 4 PWR Rods Irradiated in the CEA Osiris Reactor |
| nea-1556 |
IFPE/OVER-RAMP, Pellet Clad Interaction Failure Analysis, Power Ramps |
| nea-1776 |
IFPE/PRIMO-BD8, Belgonucleaire and SCK-CEN PRIMO Ramped MOX Fuel Rod BD8 |
| nea-1696 |
IFPE/REGATE L10.3, FGR and Fuel Swelling during power transient at medium burn-up (SILOE reactor) |
| nea-1634 |
IFPE/RISOE-1, Fission gas release from high-burnup water reactor fuel |
| nea-1502 |
IFPE/RISOE-II, Fuel Performance Data from Transient Fission Gas Release |
| nea-1493 |
IFPE/RISOE-III, Fuel Performance Data from 3rd Risoe Fission Gas Release |
| nea-1722 |
IFPE/ROPE-I, BWR, 4 rods, Ringhals, investigates clad creep-out from Studsvik (1986-1993) |
| nea-1723 |
IFPE/ROPE-II, PWR rod over pressure experiment from Studsvik |
| nea-1310 |
IFPE/SOFIT, WWER-440 Fuel Thermal Performance and Fission Gas Release |
| nea-1623 |
IFPE/SPC-RE-GINNA, Full Length and Segmented Fuel Rodlet Irradiation in PWR |
| nea-1809 |
IFPE/STEED-I Stored Energy / Enthalpy Determination from Studsvik |
| nea-1557 |
IFPE/SUPER-RAMP, PCI Failure Threshold for PWR and BWR Fuels |
| nea-1648 |
IFPE/TRANS-RAMP, Fuel behaviour data from PWR/BWR TRANS-RAMP I, II, IV experiments |
| nea-1738 |
IFPE/US-PWR-16X16 Lead Test Assembly Extended Burnup Demonstration Program |
| nea-1677 |
IFPE/ZAPOROSHYE-V1K, Zaporoshye VVER1000 fuel behaviour data (4-8 cycles, Burnup about 50 MWd/kgUO2) |
| ests0169 |
ILUCG2, 2-D Partial Differential Equations Asymmetric Matrix Solver |
| ests0170 |
ILUCG3, 3-D Partial Differential Equations Linear Asymmetric Matrix Solver |
| nesc0715 |
IMPAC-2, Dynamic Impact Analysis for 1-D Nonlinear Spring Shipping Container Model |
| ests0005 |
IMPACTS-BRC2.1, General Radiological Impacts Analysis |
| nesc0779 |
IMPORTANCE, Minimal Cut Sets and System Availability from Fault Tree Analysis |
| nesc9473 |
IMPSOR, 3-D Boundary Problems Solution for Thermal Conductivity Calculation |
| iaea1378 |
INDOSE V2.1.1, Internal Dosimetry Code Using Biokinetics Models |
| nea-0485 |
INDRA, Fusion Reactor Blanket Neutronics, Gamma Heating, H3 Breeding |
| nesc0609 |
INDX, X-Ray Diffraction Powder Pattern Indexing, Trial Unit Cell Testing |
| iaea1248 |
INDXENDF, Preparation of Visual Catalogue of ENDF Format Data |
| psr-0313 |
INFLTB, Dosimetric Mass Energy Transfer and Absorption Coefficient |
| nesc0975 |
INGEN, 2-D, 3-D Mesh Generator for Finite Elements Program |
| nesc9649 |
INGRID, 3-D Mesh Generator for Program DYNA3D and NIKE3D and FACET and TOPAZ3D |
| ccc-0185 |
INREM-EXREM-3, Time-Dependent Organ Doses from Isotope Inhalation and Ingestion |
| nea-0554 |
INSUL, Calculation of Thermal Insulation of Various Materials Immersed in He |
| nesc0590 |
INTEG INSPEC, Accident Frequencies and Safety Analysis for Nuclear Power Plant |
| nea-0744 |
INTEGR, Escape Transmission Probability in 1-D Cylindrical Geometry for Program FACEL |
| uscd1212 |
INTER, ENDF/B Thermal Cross-Sections, Resonance Integrals, G-Factors Calculation |
| nesc9480 |
INTERP, Lexical Analysis for Problems Oriented Language Development |
| iaea0886 |
INTERTRAN-I and INTERTRAN-II, Radiation Exposure from Vehicle Transport of Radioactive Material |
| uscd1213 |
INTLIB-6, Graphic Device Interface Library for ENDF/B Processing Codes |
| psr-0054 |
INTRIGUE-2L, Subroutines for Linear, Log, Semi-Log CALCOMP Plotter |
| nea-1154 |
INTRUDE, Radiation Risk from Intrusion into Shallow Land Waste Storage Site |
| nea-1153 |
INVENT, Dose Rates, Inhalation, Ingestion Risk from Closed Waste Storage Site |
| nea-1340 |
INVENT-STUDSVIK, Fission Products Abundances in U235, U238, Pu239 Samples |
| ccc-0365 |
IODES, Calculating the Estimation of Dose to the World Population from Releases of Iodine-129 to the Environment. |
| ccc-0526 |
IONMIG, Radionuclide Migration Through Porous Media |
| iaea0901 |
IPEET-103, Neutron Induced Reaction Cross-Sections for Fissile Nuclides, Preequilibrium Model |
| nea-1821 |
IPLOT, interactive MELCOR data plotting system |
| ests0109 |
IRDAM, Interactive Rapid Dose Assessment from Reactor Accident Effluents |
| nea-0513 |
IRESINT-3, Resonance Absorption in Square or Hexagonal Lattice by Single-Level Breit-Wigner |
| nea-1715 |
IRPHE-JAPAN, Reactor Physics Experiments carried out in Japan |
| iaea1415 |
IRPHE-KNK-II-ARCHIVE, KNK-II fast reactor documents, power history and measured parameters |
| nea-1660 |
IRPHE-SNEAK, KFK SNEAK Fast Reactor Experiments, Primary Documentation |
| nea-1661 |
IRPHE-ZEBRA, AEEW Fast Reactor Experiments, Primary Documentation |
| nea-1687 |
IRPHE/B&W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments |
| nea-1662 |
IRPHE/JOYO MK-II, JOYO MK-II core management and characteristics database |
| nea-1765 |
IRPHE2011-HANDBOOK, International Handbook of Evaluated Reactor Physics Benchmark Experiments |
| nea-1726 |
IRPhE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents |
| nea-1728 |
IRPhE-HTR-ARCH-01, Archive of HTR Primary Documents |
| nea-1764 |
IRPhE-TAPIRO-ARCHIVE, Fast neutron source reactor primary documents, reactor physics experiments |
| nea-1739 |
IRPhE/AVR, High Temperature Reactor Experience, Archival Documentation |
| nea-1759 |
IRPhE/BERENICE, effective delayed neutron fraction measurements |
| nea-1713 |
IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility |
| nea-1714 |
IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility |
| ests0003 |
IRRAS, Integrated Reliability and Risk Analysis System for PC |
| iaea1328 |
ISABEL EVA PACE-2, Evaporation Model with Intranuclear Cascade Input |
| nesc1034 |
ISDMS, Inel Scientific Data Management System |
| ccc-0636 |
ISO-PC, X-Ray, Gamma-Bremsstrahlung Dose-Rates |
| ccc-0079 |
ISOSHLD, Decay Gamma Dose, Bremsstrahlung Dose Behind Shield, Fission Products Source Strength |
| nea-0434 |
ISOTEX-1, Time-Dependent Heavy Isotope and Fission Products Concentration in U Reactor or Pu Reactor |
| iaea1229 |
ISOTHERM, Ion-Exchange IsoThermal Calculation and Plot |
| nesc9656 |
ITMETH, Iterative Routines for Linear System |
| ests0219 |
ITOUGH2, Inverse Modeling for TOUGH2 Multiphase Flow
Simulators |
| ccc-0467 |
ITS, TIGER System of Coupled Electron Photon Transport by Monte-Carlo |
| csni1027 |
IVO-Loop Seal Facility (Air/Water), Two-phase behaviour of a PWR cold leg loop seal during LOCA accidents |
| csni1018 |
IVO-THERMAL MIXING, study mixing of emergency cooling water with primary water during LOCA accident |
| csni1028 |
IVO/AIR-WATER-CCFL, Air/water countercurrent flow limitation experiments with full-scale fuel bundle structures |
| nesc9583 |
JAC, 2-D Finite Element Method Program for Quasi Static Mechanics Problems by Nonlinear Conjugate Gradient (CG) Method |
| iaea0940 |
JADSPE, Multi-Channel Gamma Spectra Unfolding Program |
| 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-1838 |
JASMINE V.3, Steam explosion simulation |
| nea-1811 |
JDL-IMPORTANCE, Adjoint Function: Physical Basis of Variational & Perturbation Theory in Transport & Diffusion Problems |
| nea-1843 |
JDL-REACTOR-KINETICS, Nuclear Reactor Kinetics and Control |
| nea-1844 |
JDL-THERMODYNAMICS, Thermodynamics: Frontiers and Foundations |
| nea-0317 |
JFUSER, JAERI Fast-Set Group Constant Collapsing and Data Conversion for Program LTFR-70 |
| psr-0008 |
JOMREAD, Check of 3-D Geometry Structure from Quadratic Surfaces |
| nea-0624 |
JOSHUA, Neutronics, Hydraulics, Burnup, Refuelling of LWR |
| nesc0490 |
JOSHUA-SYSTEM, Data Base Management System for Batch and Interactive Operation |
| nea-0154 |
JPHYDRO, Voids and Flow Velocity in Steady-State BWR System |
| nesc0877 |
K-FIX(3D), Transient 2 Phase Flow Hydrodynamic, X-Y-Z and Cylindrical Geometry, Eulerian Method |
| nesc0727 |
K-FIX, Transient 2 Phase Flow Hydrodynamic in 2-D Planar or Cylindrical Geometry, Eulerian Method |
| nesc0876 |
K-TIF, Thermohydraulic Dynamic of PWR in Steady-State and Transient Flow Conditions |
| nea-0492 |
KAMCCO, 3-D Time-Dependent Homogeneous and Inhomogeneous Neutron Transport by Monte-Carlo Method |
| psr-0306 |
KAOS-V, Neutron Fluence to Kerma Factor Evaluation from ENDF/B-5 and JENDL-2 |
| nea-0343 |
KASY, 3-D Homogeneous Neutron Diffusion in X-Y-Z, R-Theta, Hexagonal-Z Geometry by Synthesis Method |
| nea-1824 |
KCUT, code to generate minimal cut sets for fault trees |
| nesc0556 |
KEELE, Minimization of Nonlinear Function with Linear Constraints, Variable Metric Method |
| nea-0578 |
KEMA, KEDAK Utility, Data Update |
| ccc-0510 |
KENO-4(RG), KENO-4 with Random Geometry |
| ccc-0436 |
KENO-4/CRC, MultiGroup 3-D P1 Scattering Monte-Carlo Transport Calculation with Neutron Balance Edit |
| nea-1467 |
KENO-VA-PVM KENO-VA-SM, KENO5A for Parallel Processors |
| psr-0541 |
KENO2MCNP, Version 5L, Conversion of Input Data between KENOV.a and MCNP File Formats |
| psr-0450 |
KENO3D, Visualisation Tool for KENO V.A and KENO-VI Geometry Models |
| ccc-0548 |
KENO5A-PC, Monte-Carlo Criticality with Supergrouping |
| nea-0288 |
KERBREK, Fuel Cycle Cost Analysis for Power Reactor |
| nea-1865 |
KICHE 1.3, Kinetics of Iodine Chemistry in the Containment of LWRs under Severe Accident Conditions |
| nea-0616 |
KIM, Steady-State Transport for Fixed Source in 2-D Thermal Reactor by Monte-Carlo |
| nea-0112 |
KINAX-3, 1-D 1 Group Reactor Kinetics with Xe and I and Fission Products Heating and Auto-Control |
| nea-1002 |
KINE, 1-D PWR Dynamic with Partial Core Boiling |
| iaea1339 |
KINETIC, Time-Dependent Heat and Mass Transfer |
| nea-1293 |
KINIK, Absorber Rod Calibration Kinetics |
| nesc0528 |
KITT, Component and System Reliability Information from Kinetic Fault Tree Theory |
| ests0154 |
KIVA3, Transient Multicomponent 2-D and 3-D Reactive Flows with Fuel Sprays |
| nea-1001 |
KORIGEN, Isotope Inventory, Radiation Heat from PWR Burnup |
| nea-0417 |
KOSAK, Power Plant Cost Optimization with Pu Availability Option |
| nea-0441 |
KPD, Time-Dependent Fuel Cycle Cost Calculation for Various Reactor Types |
| ccc-0229 |
KRONIC, Annual Body Tissue Dose from Continuous Atmospheric Release |
| nesc9520 |
KRYSI, Ordinary Differential Equations Solver with Sdirk Krylov Method |
| nea-0342 |
KTOE, KEDAK to ENDF/B Format Conversion with Linear Linear Interpolation |
| nesc0987 |
L2RMAT, L**2 Method of R Matrix Propagation |
| iaea1232 |
LABAN-PEL, 2-D MultiGroup Neutron Diffusion in X-Y Geometry by Response Matrix Eigenvalue |
| nesc0992 |
LADTAP-2, Organ Doses to Man and Other Biota from Aquatic Environment |
| ccc-0696 |
LAHET 2.8, Code System for High Energy Particle Transport Calculations |
| psr-0020 |
LAPHAN0, P0 Gamma Production Matrices from ENDF/B |
| nesc0249 |
LASER, Slowing-Down Neutron Spectra and Burnup for Thermal Reactors, Neutron Transport Theory |
| nea-0573 |
LASER-PNC, Neutron Spectra in Uniform Lattice with Burnup Calculation |
| nesc0691 |
LASIP-3, CCCC Utility for BCD to BIN Conversion and BIN Data Listing |
| nesc0918 |
LASO, Subroutine Library for Matrix Manipulation, Eigenvalues and Eigenvectors |
| nea-0192 |
LAZY, General Experimental Data Processing Program |
| ests0463 |
LDEF-SS, Solve Equation Two Phase Fluid Flow in Spray Dryers |
| nea-0479 |
LEAP, Scattering Law for Continuous Phonon Spectra |
| iaea1310 |
LEGEND2010, Angular Distribution Table Calculations in ENDF Format |
| csni0004 |
LEIBSTADT/STP-2001, BWR/6, Reactor Core Isolation Cooling System Test |
| nesc0279 |
LEOPARD, Fast and Thermal Neutron Spectra from Temperature and Geometry with Depletion Calculation |
| ccc-0343 |
LEOPARD-MICRO, Spectrum-Dependent Non-Spatial Fuel Depletion |
| psr-0277 |
LEPRICON, PWR Vessel Dose Analysis with DORT and ANISN Program |
| nesc9426 |
LFK, FORTRAN Application Performance Test |
| nea-0124 |
LGH, Gamma Streaming and Neutron Streaming for Duct |
| psr-0394 |
LHS, Multivariate Sample Generator by Latin Hypercube Sampling |
| nesc1085 |
LHS-ESTSC, Multivariate Sample Generator by Latin Hypercube Sampling |
| iaea0902 |
LIANG, Neutron Induced Compound Nucleus Reaction Cross-Sections by Statistical Model |
| nea-0167 |
LIE-PN, Pn Neutron Transport in Radial Geometry Cell with Source Problems Calculation |
| nesc0460 |
LIFE-1, Stress Analysis Swelling and Performance of Cylindrical Fuel Elements in Fast Reactors |
| nea-1337 |
LIMES, IMF in Heavy Ion Nuclear Reaction by Sum-Rule Model |
| nesc0657 |
LINDA, Diagnostics of Stress Analysis of Linear Elastic Structure by Least Square Fit |
| iaea1311 |
LINEAR2010, Linear-Linear Interpolation of ENDF Format Cross-Sections |
| nesc0800 |
LINPACK, Subroutine Library for Linear Equation System Solution and Matrix Calculation |
| iaea1331 |
LINTAB, Linear Interpolable Tables from any Continuous Variable Function |
| psr-0117 |
LINX, MINX Library Utility, Data Merge |
| nea-0860 |
LISA, Hazard Assessment of Nuclear Waste Disposal in Geological Formations |
| uscd1214 |
LISTEF, ENDF/B Data File Summary List |
| nesc0638 |
LISTF-4, ENDF/B Utility, Data Listing |
| csni0034 |
LOBI/A1-04R, Loop for Blowdown Investigation, PWR Double-Ended Cold-Leg Break Test |
| csni0035 |
LOBI/A1-06, Loop for Blowdown Investigation, PWR Double-Ended 2A Cold-Leg Break |
| csni0036 |
LOBI/A1-66, Loop for Blowdown Investigation, PWR Double-Ended 2A Cold-Leg Break |
| csni0037 |
LOBI/A2-77, Loop for Blowdown Investigation, PWR MOD2 Small Leak Programme Experiment |
| csni0038 |
LOBI/A2-81, Loop for Blowdown Investigation. PWR MOD2 1% Cold-Leg Break |
| csni0003 |
LOBI/B-R1M, Loop for Blowdown Investigation, PWR Single-Ended Cold-Leg Break Experiment B |
| csni0074 |
LOBI/BT-00, Simulation of a Loss of Feedwater Transient (LOFW) |
| nea-0623 |
LOCA-MARK-2, Fuel Temperature and Clad Temperature in HWR Steam Generator LOCA |
| csni0017 |
LOFT/L2-3, Loss of Fluid Test, 2nd NRC L2 Large Break LOCA Experiment |
| csni0016 |
LOFT/L2-5, Loss of Fluid Test, 3rd NRC L2 Large Break LOCA Experiment |
| csni0022 |
LOFT/L3-5, Loss of Fluid Test, 5th NRC L3 Small Break LOCA Experiment |
| csni0018 |
LOFT/L3-6, Loss of Fluid Test, 6th NRC L3 Small Break LOCA Experiment |
| csni0021 |
LOFT/L3-7, Loss of Fluid Test, 7th NRC L3 Small Break LOCA Experiment |
| csni0020 |
LOFT/L6-7, Loss of Fluid Test, Anticipated Transients with Multiple Failures |
| csni0070 |
LOFT/L8-2, Severe Core Transient Experiment |
| csni0019 |
LOFT/L9-3, Loss of Fluid Test, Anticipated Transients with Multiple Failures |
| csni0010 |
LOFT/LP-02-6, Loss of Fluid Test, 1st OECD Large Break Experiment |
| csni0012 |
LOFT/LP-FP-1B, Loss of Fluid Test, Fission Product Release Experiment |
| csni0013 |
LOFT/LP-FP-2, Loss of Fluid Test, Fission Product Release from Fuel |
| csni0007 |
LOFT/LP-FW-1, Loss of Fluid Test, PWR Response to Loss-of-Feedwater Transient |
| csni0002 |
LOFT/LP-LB-1, Loss of Fluid Test, Large-Break LOCA Experiment |
| csni0008 |
LOFT/LP-SB-1, Loss of Fluid Test, Small Hot Leg Break LOCA, Early Pump |
| csni0009 |
LOFT/LP-SB-2, Loss of Fluid Test, Small Hot Leg Break LOCA, Delayed Pump |
| csni0011 |
LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressureinjection System (HPIS) |
| nea-0965 |
LOLA-SYSTEM, JEN-UPM PWR Fuel Management System Burnup Code System |
| nea-0185 |
LOOP-3, Hydraulic Stability in Heated Parallel Channels |
| nea-1026 |
LOUHI, Generator Spectra Unfolding Program with Linear and Nonlinear Regularization |
| iaea1304 |
LPA1, LPA2, Deconvolution Program Using Fourier Transform |
| nesc9449 |
LPGC, Levelized Steam Electric Power Generator Cost |
| ccc-0385 |
LPGS, Radiation Exposure from Radioactive Release into Hydrosphere |
| ccc-0064 |
LPSC, Protons and Neutron Flux, Spectra Behind Slab Shield from Protons Irradiation |
| iaea1260 |
LPTAU, Quasi Random Sequence Generator |
| nesc9721 |
LRSYS, PASCAL LR(1) Parser Generator System |
| nesc1033 |
LSAP-DIGLIB, Linear Control System Design, Analysis, Plotting |
| nea-1306 |
LSHINSE, Air Scattering Neutron and Gamma Doserates for Complex Shielding Geometry |
| psr-0233 |
LSL-M2, Neutron Spectra Log Adjustment for Dosimetry Applications |
| uscd1227 |
LSODA, Ordinary Differential Equation Solver for Stiff or Non-Stiff System |
| uscd1228 |
LSODAR, Ordinary Differential Equation Solver for Stiff or Non-Stiff System with rootfinding |
| uscd1223 |
LSODE, 1st Order Stiff or Non-Stiff Ordinary Differential Equations System Initial Value Problems |
| uscd1229 |
LSODES, Ordinary Differential Equations System Sparse Matrices |
| uscd1224 |
LSODI, Implicit Ordinary Differential Equations System Either Dense or Banded Matrices |
| uscd1225 |
LSODIS, Implicit Ordinary Differential Equations System Sparse Matrices |
| ests0264 |
LSODKR, Stiff Ordinary Differential Equations (ODE) System Solver with Krylov Iteration and Rootfinding |
| uscd1230 |
LSODKR, Stiff Ordinary Differential Equations (ODE) System Solver with Krylov Iteration with Rootfinding |
| uscd1231 |
LSODPK, Ordinary Differential Equations Solver for Stiff and Nonstiff System with Krylov Corrector Iteration |
| uscd1226 |
LSOIBT, Implicit Ordinary Differential Equations System Block Tridiagonal Matrices |
| iaea1268 |
LSQXY, Curve Fitting with Uncertainty Weighting |
| nea-0316 |
LTFR-4, Library Generated for Fast Reactor Design Program from JAERI Fast-Set MultiGroup Constant |
| nesc0648 |
LUGS, Stress Analysis, Flexibility Factors for Rectangular Attachment on Thin Shell |
| ccc-0220 |
LUIN-II, Cosmic Ray Cascade Generator and Particle Fluxes |
| nea-0250 |
LUPO, Temperature and Void Rate and Pressure Drop and Flow Rate in Pressure Loop |
| ccc-0631 |
LWRARC, PWR and BWR Spent Fuel Decay Heat Generator |
| nesc0381 |
LYNNE, Inelastic Scattering by Multipole Expansion of Woods-Saxon |
| psr-0132 |
MACK, Fluence to Kerma Generator from ENDF/B |
| nesc0574 |
MACS, Lattice Vibrations Structure Factors for Thermal Neutron Scattering in Moderators |
| nea-0836 |
MADONNA, Neutron Flux with Void Region by Removal Diffusion Method |
| nesc1006 |
MAEROS, Multicomponent Aerosol Time Evolution |
| ccc-0359 |
MAGIK, Photon Dose Rates from Nucleon-Nucleus % Meson-Nucleus Collisions |
| ests0386 |
MAGNUM-2D, Heat Transport and Groundwater Flow in Fractured Porous Media |
| nea-0931 |
MAIA, Eigenvalues for MHD Equation of Tokamak Plasma Stability Problems |
| nea-0565 |
MAILLE, Triangular Finite Elements Generator for Planar Structure |
| nesc0256 |
MANTA, Heat Transfer Fuel Elements Cluster to Single-Phase Steady-State Fluid Flow |
| nea-1047 |
MANYCASK, Radiation Dose Rate Around Many Casks |
| nea-1096 |
MAPLE, Fault Tree Plotting |
| nea-0517 |
MAPLIB, Thermodynamics Materials Property Generator for FORTRAN Program |
| nesc0939 |
MAPPER, Graphics for Transparencies and Slides Using DISSPLA System |
| nea-0528 |
MARC, MultiGroup Diffusion in R-Z, X-Y, R-Theta, Slab, Cylindrical, Spherical, Hexagonal, Triangular Geometry |
| nesc0734 |
MARCH, Containment Behaviour after LOCA, Blowdown, Meltdown, Metal H2O Reaction |
| nea-1017 |
MARCOPOLO, Radial and Axial Diffusion Coefficient for Cylindrical Wigner-Seitz Reactor Cell |
| nea-0526 |
MARE, Reaction Cross-Sections by Blatt-Ewing Statistical Evaporation Model |
| nea-0926 |
MARIA-SYSTEM, PWR Fuel Assembly Calculation for Program CARMEN-System and Simulation |
| ccc-0503 |
MARINRAD, Health Hazard from Radioactive Material Release into Ocean |
| psr-0137 |
MARLOWE 15b, Computer Simulation of Atomic Collisions in Crystalline Solids |
| nea-1307 |
MARMER, Point-Kernel Shielding Calculation with Nuclide Concentrations from ORIGEN-S |
| psr-0117 |
MARS-ORNL, Processing Program Collection for AMPX, CCCC, ANISN, DOT, MORSE Format Library |
| nea-0983 |
MARTHA, Nai(Tl) Gamma Scintillation Detector Response by Monte-Carlo |
| 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-XTREME, 1 Constraint Lagrange Multipliers for 25 Variables |
| ests0221 |
MAXWELL3, 3-D FEM Electromagnetics |
| nesc9907 |
MAZE, Input Generator for Program DYNA2D and NIKE2D |
| nesc0355 |
MC**2-2, MultiGroup Neutron Spectra, Slowing-Down Calculation Using ENDF/B, P1 and B1 Approximation |
| psr-0350 |
MC*2-2, Calculation of Fast Neutron Spectra and Multigroup Cross-Sections from ENDF/B Data |
| nea-1643 |
MCB1C, Monte-Carlo Continuous Energy Burnup Code |
| csni2003 |
MCCI PROJECT, Molten Core Concrete Interaction Project |
| csni2017 |
MCCI-2 PROJECT, Melt Coolability and Concrete Interaction Phase 2 Project |
| nea-0452 |
MCDATA, MC**2 Cross-Sections Conversion for Programs CITATION, ANISN, DOT |
| nea-1632 |
MCDSIM, Atmospheric Monte Carlo Dispersion Simulation |
| nea-1733 |
MCNP4B-GN, Monte Carlo Code System for (gamma,n) production and transport in high-Z materials |
| iaea0889 |
MCRAC, In Core Fuel Management, Program of PFMP System |
| nea-0971 |
MCRTOF, Multiple Scattering of Resonance Region Neutron in Time of Flight Experiments |
| ests1678 |
MCSLTT, Monte Carlo Simulation of Light Transport in Tissue |
| nea-1166 |
MCVIEW, 3-D Radiation View Factor by Monte-Carlo Method |
| ccc-0156 |
MECC-7, Medium-Energy Intranuclear Cascade Code System |
| nea-0362 |
MEDEA, Steady-State Pressure and Temperature Distribution in He H2O Steam Generator |
| nea-1140 |
MEDUSA-1B, 1-D Plasma Hydrodynamic Analysis of Fusion Pellet Ion Beams |
| nea-0583 |
MEDUSA-PIJ, 1-D Thermohydraulic Analysis of Laser Driven Plasma |
| nea-1057 |
MELODIE, Radiological Assessment of Nuclear Waste Migration in Ground Water |
| nesc0700 |
MELT-3, Thermohydraulics and Neutronics, Fast Reactor Transients with Feedback |
| nea-0351 |
MERCURE, 3-D Gamma Heating and Gamma Dose Rate and Fast Flux by Monte-Carlo |
| nea-0194 |
MERCURE-3, Gamma Attenuation by Line-of-Flight in 3-D Heterogeneous Geometry |
| iaea1312 |
MERGER2010, Merges ENDF/B Data by Material Number or Identifier |
| nesc0825 |
MESA, Fourier Analysis of Maximum Entropy Spectra and Correlation Function Analysis |
| nea-0346 |
MESHGEN, Triangular Finite Elements Generator |
| nea-0348 |
MESHPLOT, CALCOMP Plot of 2-D Triangular Finite Elements Mesh |
| nea-0347 |
MESHREF, Finite Elements Mesh Combination with Renumbering |
| nesc9862 |
MESOI2.0, Atmospheric Transport of Effluent Puffs |
| ests0331 |
MESORAD, Emergency Response Airborne Dose Assessment |
| nea-1534 |
MESYST, Simulation of 3-D Tracer Dispersion in Atmosphere |
| iaea1387 |
MEXP, EXTERMINATOR-2 Utility Programs |
| nesc9479 |
MGA, Pu Isotope Abundance from Multichannel Analyzer Gamma Spectra |
| psr-0542 |
MGA8, Determine Pu Isotope Abundances from Multichannel Analyzer Gamma Spectra |
| ests0233 |
MGMHD, Multigrid 3-D for the Analysis of Magnetohydrodynamic (MHD) Channels |
| psr-0261 |
MICAP, Ionization Chamber Detector Response by Monte-Carlo |
| nea-1562 |
MICROX-2, Group Constant Generator with Resonance Interference and Self-Shielding |
| nea-0388 |
MIGROS, Group Cross-Sections, Self-Shielding Factors, Scattering Transfer, Fission from KEDAK |
| nesc9460 |
MILDOS-AREA, Radiological Impact of Airborne U238 from Mining and Milling |
| uscd1097 |
MINEQL, Chemical Equilibrium Composition of Aqueous Systems |
| ests0143 |
MINET, Transient Fluid Flow and Heat Transfer Power Plant Network Analysis |
| nea-0639 |
MINIGAL, Average Thermal Cross-Sections, Epithermal Cross-Sections, Fission Cross-Sections from UKNDL |
| nesc0888 |
MINPACK-1, Subroutine Library for Nonlinear Equation System |
| nesc1101 |
MINTEQ, Geochemical Equilibria in Ground Water |
| psr-0105 |
MINX, MultiGroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX |
| nea-0474 |
MISSIONARY, ENDF/B to UKNDL Format Conversion |
| iaea1313 |
MIXER2010, Cross Sections Calculations for a Composite Mixture of ENDF Format Material |
| nesc0632 |
MMM-3, Semi Rigid Molecule Normal Modes and Frequencies for Slow Neutron Scattering Calculation |
| nea-1706 |
MMRW, Canadian and early British Energy Reports on Nuclear Reactor Theory (1940-1946) |
| nea-1792 |
MMRW-BOOKS, Legacy books on slowing down, thermalization, particle transport theory, random processes in reactors |
| nesc9853 |
MMT, 1-D Radionuclide Groundwater Transport |
| nea-1005 |
MOBIDIC, Fast Reactor Hexagonal Infinite Lattice 2 Component Fuel Pin Diffusion Coefficient |
| iaea1238 |
MOCA, Criticality of VVER Reactor Hexagonal Fuel Assemblies |
| psr-0365 |
MOCUP, MCNP/ORIGEN Coupling Utility Programs |
| nesc0653 |
MOCUS, Minimal Cut Sets and Minimal Path Sets from Fault Tree Analysis |
| nesc0491 |
MOD-5, Time-Dependent MultiGroup Slowing-Down Neutron Spectra and Keff Calculation, Green Function Method |
| nea-0540 |
MODESTY, Statistical Reaction Cross-Sections and Particle Spectra in Decay Chain |
| nea-1279 |
MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors |
| nea-1762 |
MODLIB, library of Fortran modules for nuclear reaction codes |
| nea-1414 |
MOLGEO, Molecular Structure Data Tables |
| nea-0527 |
MONK, Keff, Collision Rate, Flux Distribution in General Geometry from UKNDL by Monte-Carlo Method |
| nea-1747 |
MONTE-CARLO-WS-2005, Proceedings of Monte Carlo Criticality Calculations & TRIPOLI-IV Workshops 2005 |
| psr-0455 |
MONTEBURNS 2.0: An Automated, Multi-Step Monte Carlo Burnup Code System |
| psr-0411 |
MORECA, Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup |
| ccc-0127 |
MORSE, MultiGroup Neutron Transport and Gamma Transport for Complex Geometry Shields by Monte-Carlo |
| ccc-0431 |
MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method |
| ccc-0474 |
MORSE-CGA, Monte-Carlo Neutron and Gamma MultiGroup Transport with Array Geometry |
| nea-1181 |
MORSE-DD, Monte-Carlo Neutron Transport with Combinatorial Geometry Using DDL Cross-Sections Library |
| ccc-0127 |
MORSE-E, Program MORSE with Uniform Source for Various Geometry |
| ccc-0588 |
MORSE-EMP, Monte-Carlo Neutron and Gamma MultiGroup Transport with Array Geometry, for PC |
| psr-0142 |
MORSEC-SP, Step Function Angular Distribution for Cross-Sections Calculation by Program MORSE |
| nesc0678 |
MORTRAN-2, FORTRAN Language Extension with User-Supplied Macros |
| nea-1633 |
MOSRA-LIGHT, High Speed 3-D X-Y-Z Nodal Diffusion Code for Vector Computers |
| nesc0551 |
MOXY-MOD32, Thermal Analysis Swelling and Rupture of BWR Fuel Elements During LOCA |
| nesc0551 |
MOXY/MOD-1, Thermal Analysis Swelling and Rupture of BWR Fuel Elements During LOCA |
| ests1098 |
MPICH, Message Passing Interface (MPI) Subroutine Library for Parallel Computers and Networks |
| nesc0798 |
MSF21/VTE21, Desalination Plant Heat, Mass Balance, Design, Cost Optimization |
| nesc0508 |
MUCHA1, Fuel Rod Pair Thermohydraulics During LOCA and ECCSA for LWR |
| nesc0508 |
MUCHA2, Primary Coolant Thermohydraulics During LOCA and ECCS for LWR |
| nea-0816 |
MUENSTER, 2-D R-Z Geometry Thermohydraulics Calculation for Pebble-Bed Reactor |
| nea-0933 |
MULTI-KENO, Criticality Safety Analysis by Monte-Carlo |
| nea-1041 |
MULTIPLET, Large Event Trees for Risk Assessment Calculation |
| nesc9684 |
MULTITASKER, Multitasking Kernel for C and FORTRAN Under UNIX |
| iaea0907 |
MUP-2, Fast Neutron Nuclear Reaction Cross-Sections of Medium-Heavy Nuclei |
| nea-0035 |
MUPO, Critical 43 Group Spectra Calculation for Homogeneous Reactor |
| nea-1845 |
MURE, MCNP Utility for Reactor Evolution: couples Monte-Carlo transport with fuel burnup calculations |
| iaea0890 |
MURLI, 1-D Flux, Reaction Rate in Cylindrical Geometry Thermal Reactor Lattice by Transport |
| iaea0892 |
MURLI-CLUSTER, Lattice Calculation of PHWR with Rod Clustered Fuel |
| nea-1451 |
MUTIL, Asymmetry Factor of Mott Cross-Sections for Electron, Positron Scattering |
| nea-1673 |
MVP/GMVP II, MC Codes for Neutron & Photon Transport Calc. based on Continuous Energy and Multigroup Methods |
| iaea1411 |
NAAPRO, Neutron Activation Analysis Prognosis and Optimization code |
| ccc-0164 |
NAC, Neutron Activation Analysis and Isotope Inventory |
| nesc9489 |
NACHOS2, Incompressible Viscous Fluid Dynamic |
| nesc0717 |
NAHAMMER, Pressure Transients in Na LMFBR Piping System, Linear Fluid Hammer Theory |
| nea-0806 |
NAIAD, LOCA Transient and Steady-State 2 Phase Flow in Channel Network |
| psr-0085 |
NAISAP, Theory and Use of Gamma-Ray Spectrum Analysis Codes for NaI(Tl) Detectors. |
| nesc0780 |
NALAP, Thermohydraulics for Na Cooled LMFBR after Pipe Rupture and Accidents |
| iaea0863 |
NANICK, Infinitely Dilute Group Constant and Scattering Matrix from ENDF/B |
| nesc9644 |
NASA-VOF2D, 2-D Transient Free Surface Incompressible Fluid Dynamic |
| nesc9568 |
NASA-VOF3D, 3-D Transient, Free Surface, Incompressible Fluid Dynamic |
| nesc0719 |
NATRAN-2, LMFBR Piping System Pressure Transients, Fluid Hammer and Na H2O Reaction |
| nesc0718 |
NATRANSIENT, LMFBR Piping System Pressure Transients, Fluid Hammer, Na H2O Reaction |
| nea-0853 |
NAUA-MOD5 NAUA-MOD5/M, Aerosols in Reactor Containment During Meltdown |
| ccc-0462 |
NCRP49, X-Ray Shielding for Radiographic and Fluoroscopic Diagnostic Units |
| nea-0599 |
NE-SPEC, Ne-213 Liquid Scintillation Detector Fast Neutron Spectra Unfolding |
| 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 2.3.16, Evaluation of Routine Radioactive Effluents from Nuclear Power Plants |
| ccc-0768 |
NRCDOSE72 1.2.1, Evaluation of Routine Radioactive Effluents from Nuclear Power Plants with Windows Interface |
| ests1049 |
NRCPIPES, Fracture Mechanics of Cracked Pipes |
| nea-0700 |
NRESP-3, Organic Scintillation Detector Response to Monoenergetic Fast Neutron |
| nea-0125 |
NRN, Removal-Diffusion for Squares and Cylindrical Geometry with Energy Transfer Matrix |
| iaea1389 |
NRSC, Neutron Resonance Spectrum Calculation System |
| nea-1347 |
NSLINK, Coupling of NJOY Cross-Sections Generator Code to SCALE-3 System |
| nesc0790 |
NUBOW-2D/INEL, 2-D Core Restraint System Stress Analysis, with Bowing, Creep, Swelling |
| nea-0951 |
NUCCON, Nuclide Concentration and Activation in D-T Fusion Reactor |
| iaea1320 |
NUCHART, Nuclear Properties and Decay Data Chart |
| nea-1492 |
NUCLEUS-CHART, Interactive Chart of Nuclides |
| nesc0683 |
NUFUEL, Conditions for Power Production, U Fuel, Pu Recycle and Reprocessing |
| nesc9888 |
NUTRAN, Doses by Radionuclide Migration from Nuclear Waste Storage |
| iaea0918 |
NX-1, Excitation Function of (N-P) and (N-He4) Reaction |
| iaea0919 |
NX-2, Excitation Function of (N-D) and (N-He3) Reaction |
| psr-0014 |
O5S, Calibration of Organic Scintillation Detector by Monte-Carlo |
| nesc1125 |
OCA-P, PWR Vessel Probabilistic Fracture Mechanics |
| nesc0753 |
OCOPTR, Minimization of Nonlinear Function, Variable Metric Method, Derivative Calculation |
| nesc0898 |
OCTAVIA, PWR Pressure Vessel Failure Probability for Routine Pressure Transients |
| uscd1232 |
ODEPACK, Initial Value Problems of Ordinary Differential Equation System |
| ccc-0046 |
OGRE, Monte-Carlo System for Gamma Transport Problems |
| csni2014 |
OLHF, Sandia Lower Head Failure of the reactor pressure vessel OECD/NEA Project |
| nea-1591 |
OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo |
| nea-1271 |
OMICRON, LLNL ENDL Charged Particle Data Library Processing |
| ccc-0266 |
ONETRAN, 1-D Transport in Planar, Cylindrical, Spherical Geom. for Homogeneous, Inhomogeneous Probl., Anisotropic Source |
| nea-0552 |
OPTIM, Minimization of Band-Width of Finite Elements Problems |
| nesc0829 |
OPTIMIZERS, Subroutine Library for Unconstrained Nonlinear Optimization Problems |
| iaea1316 |
OPTMOD, Elastic and Total Cross-Sections, Polarization by Optical Model |
| nesc0703 |
ORCENT-2, Full Load Steam Turbine Cycle Thermodynamics for LWR Power Plant |
| nesc0588 |
ORCOST-2, PWR, BWR, HTGR, Fossil Fuel Power Plant Cost and Economics |
| nea-1324 |
OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN |
| ccc-0371 |
ORIGEN-2.2, Isotope Generation and Depletion Code Matrix Exponential Method |
| ccc-0702 |
ORIGEN-ARP 2.00, Isotope Generation and Depletion Code System-Matrix Exponential Method with GUI and Graphics Capability |
| nea-0622 |
ORIGEN-JR, Radiation Source and Nuclide Transmutation with In-Core Burnup |
| nesc9906 |
ORION, Postprocessor for Finite Elements Program NIKE2D and DYNA2D |
| nea-1249 |
ORION-II, Concentration and Dose from Radioactive Release into Atmosphere |
| ccc-0731 |
ORIP-XXI, isotope transmutation simulations |
| ests0329 |
ORMGEN3D, 3-D Crack Geometry FEM Mesh Generator |
| psr-0275 |
ORMONTE, Uncertainty Analysis for User-Developed System Models |
| nesc0699 |
ORSIM, Nuclear Fuel, Fossil Fuel Hydroelectric Power Plant Cost and Economics |
| nesc0525 |
ORTHAT, Transient Heat Conduction in 2-D X-Y, R-Z and R-Theta Geometry |
| nesc0525 |
ORTHIS, Steady-State Heat Conduction in 2-D X-Y, R-Z and R-Theta Geometry |
| nesc1102 |
ORTURB, HTGR Steam Turbine Dynamic for FSV Reactor |
| csni0014 |
OTIS/220100, Once-Through Integral Systems, 2-Phase Natural Circulation and Reflux |
| csni0015 |
OTIS/220402, Once-Through Integral Systems, Cold Leg Small Break LOCA |
| nesc9469 |
OTTER, Resolution Style Theorem Prover |
| nea-0802 |
OWEN-1, LOCA Transient and Steady-State 2 Phase Flow in Heated Channel |
| psr-0538 |
P-CARES 2.0.0, Probabilistic Computer Analysis for Rapid Evaluation of Structures |
| nesc0926 |
PABLM, Doses from Radioactive Releases to Atmosphere and Food Chain |
| nesc0540 |
PACTOLUS, Nuclear Power Plant Cost and Economics by Discounted Cash Flow Method |
| nesc0901 |
PAD, Coupled Neutronics, Thermohydraulics in 1-D Spherical, Cylindrical, Planar Geometry |
| ccc-0621 |
PAGAN-1.1, Low-Level Nuclear Waste in Ground Water, Performance Assessment Code |
| csni2004 |
PAKS PROJECT, the fuel behaviour in accident conditions on the basis of analyses of the PAKS-2 event |
| nea-1008 |
PALLAS-1D(VII), Direct Integration of Transport Equation in 1-D Planar and Spherical Geometry |
| nea-0702 |
PALLAS-2DY, 2-D Neutron Transport, Gamma Transport with Anisotropic Scattering for Fixed Source |
| psr-0156 |
PAPIN, Cross Section, Self-Shielding Factors for Fertile Isotopes in Unresolved Resonance Region |
| nesc0555 |
PARET-ANL(NESC), Thermohydraulics of Reactivity Accident in LWR |
| psr-0516 |
PARET-ANL, Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores |
| ccc-0499 |
PART61, Low Level Radioactive Waste Impact Analysis |
| ccc-0760 |
PARTISN 5.97, 1-D, 2-D, 3-D Time-Dependent, Multigroup Deterministic Parallel Neutral Particle Transport Code |
| nea-0521 |
PAS-1, 2-D, 3-D Linear Static and Dynamic Stress Analysis with 2-D Steady-State Temperature Distribution |
| nea-1238 |
PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation |
| nea-1680 |
PASCAL, Probabilistic Fracture Mechanics Analysis of Structural Components in Aging LWR |
| csni1014 |
PATRICIA/GV-6, Steady State Steam Generator Test Facility |
| nesc9797 |
PATTER, Pattern Recognition Data Analysis |
| ccc-0445 |
PAVAN, Atmospheric Dispersion of Radioactive Releases from Nuclear Power Plants |
| nesc9617 |
PC-BLAS, PC Linear Algebra Subroutines |
| ests0071 |
PC-PRAISE, BWR Piping Reliability Analysis |
| nesc1057 |
PCC/SRC, PCC and SRC Calculation from Multivariate Input for Sensitivity Analysis |
| ests0764 |
PCDOSE-ESTSC, Radioactive Dose Assessment and NRC Verification |
| nesc9917 |
PCHIP, Piecewise Cubic Hermite Data Interpolation |
| uscd1205 |
PCNUDAT-PCNULIB, Nuclear Properties Data Base and Retrieval System |
| iaea1220 |
PCROSS, Pre-Equilibrium Emission Spectra in Neutron Reactions |
| ests1145 |
PCX, Interior-Point Linear Programming Solver |
| ests0847 |
PDASAC, Partial Differential Sensitivity Analysis of Stiff System |
| nesc9839 |
PDES, Fips Standard Data Encryption Algorithm |
| csni1002 |
PDHT-HP, Post Dryout Heat Transfer Experiments, Upflow and Downflow Conditions |
| csni1003 |
PDHT-LP, Low Pressure Post Dryout Loop, Upflow Conditions |
| iaea1261 |
PEGAS, Unified Model for Particle and Gamma Emission Nuclear Reactions |
| nesc0865 |
PELE-IC, 2-D Eulerian Incompressible Hydrodynamic and Bubble Dynamic after LWR LOCA |
| iaea0819 |
PELINOMIC, Power Plant Cost Optimization for Dispersed Load Centres |
| iaea0829 |
PELINSCA, Elastic Scattering and Total Cross-Sections and Polarization by Hauser-Feshbach |
| iaea0855 |
PELSHIE, Dose Rates from Gamma Source by Point-Kernel Integration |
| nea-1525 |
PENELOPE2011, A Code System for Monte-Carlo Simulation of Electron and Photon Transport |
| nea-1339 |
PEPIN, Methodology for Computing Concentrations, Activities, Gamma-Ray Spectra, and Residual Heat from Fission Products. |
| iaea1185 |
PEQAG-2, Pre-Equilibrium Model Nucleon, Gamma Spectra and Cross-Sections |
| nesc9800 |
PFPL, Puff Plume Atmospheric Radioactive or Toxic Deposition |
| iaea1413 |
PGAA-IAEA, Database for Prompt Gamma-ray Neutron Activation Analysis |
| uscd1222 |
PHAST, Calculation of isotope equilibrium constants for geochemical models |
| psr-0432 |
PHAZE, Parametric Hazard Function Estimation |
| csni1025 |
PHEBUS/B9+ Degradation of a PWR Type Core during a severe fuel damage |
| csni1021 |
PHEBUS/test-218, Behaviour of a Fuel Rod Bundle during a Large Break LOCA Transient with a two Peaks Temperature History |
| nesc0454 |
PHENIX, 2-D MultiGroup Diffusion Fast Reactor Burnup Calculation and Fuel Cycle Analysis |
| nea-1857 |
PHITS-2.24, Particle and Heavy Ion Transport code System |
| uscd1207 |
PHREEQC, Modeling of Geochemical Reactions, Calculation of pH, REDOX Potential |
| uscd1207 |
PHREEQCI, Windows Interactive Version of PHREEQC |
| nesc9674 |
PHREEQE, Modelling of Geochemical Reaction, Calculation of P-H, Redox Potential |
| uscd1207 |
PHRQCGRF, code to create graphs from the data generated by PHREEQC |
| uscd1183 |
PHRQPITZ, Geochemical Calculation in Brines |
| ccc-0160 |
PICA, Photon-Induced Medium-Range Nuclear Cascade Analysis by Monte-Carlo |
| psr-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-REPORTS, Final reports on the PMK-2 projects for VVER Safety Studies |
| nea-0464 |
PN, 1-D, 2-D, 3-D MultiGroup Neutron Transport |
| ests0428 |
POISSON SUPERFISH, Poisson Equation Solver for Radio Frequency Cavity |
| nea-1058 |
POISSON, Analysis Solution of Poisson Problems in Probabilistic Risk Assessment |
| nea-0488 |
POISSX, Poisson Equation on Rectangle with Various Boundary Condition |
| nesc0639 |
POLLA-NESC, Resonance Parameter R-Matrix to S-Matrix Conversion by Reich-Moore Method |
| iaea0944 |
POLLA/IECTA, ENDF/B Reich-Moore to Adler-Adler Resonance Parameter Conversion |
| nesc9680 |
POSSOL, 2-D Poisson Equation Solver for Nonuniform Grid |
| iaea1249 |
POTAUS, Stopping Power and Particle Ranges in Various Material |
| nesc0340 |
POWERCO, Nuclear Power Plant Electricity Cost and Economics |
| nea-1675 |
PPICA, Power Plant Investment Cost Analysis |
| nesc1070 |
PRAISE-C, Double-Ended Guillotine Break (DEGB) Breaks from Weld Cracks in Light-Water Reactor Piping System |
| nesc9983 |
PRAXIS, High Level Computer Language for System Applications |
| nea-0809 |
PREANG, Spectra and Angular Distribution from Nuclear Reaction by Statistical Model |
| nea-0904 |
PRECIP-2, Zircaloy Cladding Oxidation Simulation for LWR under LOCA Conditions |
| psr-0226 |
PRECO-2000, Exciton Model Preequilibrium Code System with Direct Reactions |
| psr-0226 |
PRECO-D2, Pre-Equilibrium and Direct Reaction Double Differential Cross-Sections |
| nea-0509 |
PREDEX-1, U, Pu, Nitric Acid Distribution in Counter Current Solvent Extraction |
| nea-0888 |
PREM, Pre-Equilibrium Energy Spectra and Cross-Sections for Multiple Nucleon Emission |
| nesc0528 |
PREP KITT, System Reliability by Fault Tree Analysis |
| nea-1173 |
PREP, Input Preparation for Monte-Carlo Program SPOP |
| nesc0528 |
PREP, Min Path Set and Min Cut Set for Fault Tree Analysis, Monte-Carlo Method |
| nea-1485 |
PREP-45, Input Preparation for CITATION-2 |
| iaea1379 |
PREPRO2010, Data Preparation and Management, Subsidiary Calculations (ENDF Format) |
| nea-0251 |
PREST, Pressure Temperature Transients, I Inhalation in Containment Building from LOCA |
| iaea0905 |
PRESTO, Slab Shields for Time-Dependent Gamma Spectra |
| ccc-0504 |
PRESTO-II, Low Level Radioactive Waste Transport and Risk Assessment |
| iaea0817 |
PROB, Transport Equation in Slab Geometry and Collision Probability by Overrelaxation Method |
| nea-0695 |
PROCIV, Protection Coefficient from Fallout in Residential Area Housing |
| nea-0169 |
PROCOPE, Collision Probability in Pin Clusters and Infinite Rod Lattices |
| nea-1170 |
PROF-DD, Generator of MultiGroup Cross-Sections Library DDX for MORSE-DD, ANISN-DD, DOT-DD |
| nesc1023 |
PROGRAM-H, Analysis of Transonic Airfoils with Turbulent Boundary Layer Correlation |
| ests0790 |
PROGRAM-K, Transonic Airfoil, Turbine, Compressor Blade Design |
| nesc0846 |
PROMSYS, Plant Equipment Maintenance and Inspection Scheduling |
| iaea1216 |
PRORIA, Fast Reactivity Transients in PWR with Two-Phase Flow Model |
| nesc0778 |
PROSA-1 PROSA-2, Accidents Probability Analysis Using Response Surface Method |
| nesc0542 |
PSA-2, Stress Analysis, Thermal Expansion and Loads in Multi Anchor Piping System |
| nea-1138 |
PSACON, Conversion Program for PSAOUT-I Output Files |
| iaea1174 |
PSAPACK, Probabilistic Safety Analysis with Fault Event Trees |
| csni2200 |
PSB-VVER, Computer code validation for transient analysis of VVER and RBMK reactors project |
| iaea0888 |
PSU-LEOPARD, Program LEOPARD in PFMP System, Fast Neutron and Thermal Neutron Spectra Calculation |
| uscd1216 |
PSYCHE, ENDF/B Data Consistency Check in ENDF Format |
| nesc0155 |
PTH-1, Pressure and Temperature in Containment after Blowdown of H2O Coolant System |
| ccc-0618 |
PTRAN, Proton Transport for 50 to 250 MeV by Monte-Carlo |
| psr-0157 |
PUFF-2, MultiGroup Covariance Matrices from ENDF/B-5 Error Files |
| psr-0534 |
PUFF-IV 6.1.0, Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files |
| iaea1228 |
PULSTRI, Mixed Core Triga Reactor Pulse Calculation |
| ccc-0595 |
PUTZ, Point-Kernel 3-D Gamma Shielding |
| nea-1679 |
PVIS-4, Pressure vessel irradiation, source preparation |
| nesc0441 |
PWCOST, Fuel Cycle Cost and Economics by Present Worth Levelized Method |
| nesc1081 |
PWR-GALE, Radioactive Gaseous Release and Liquid Release from PWR |
| nea-1780 |
PWR-MOX/UOX-TRANS, OECD/NEA US-NRC PWR MOX/UO2 Core Transient Benchmark |
| nesc0552 |
PWR-PPM, Boration-Dilution Tables Generator for PWR Operation |
| 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-0083 |
RAID, Gamma, Neutron Scattering into Cylindrical or Multibend Duct |
| iaea0822 |
RAM-1, Thermal Flux Derivatives at Plane Geometry Control Rod Boundary by Monte-Carlo |
| nea-0843 |
RANCH, Radionuclide Migration in Geological Media |
| nea-0939 |
RANDOM, Random Number Generator with Large Cycle Length |
| nesc0843 |
RANDOM_NUMBERS, Random Number Sequence Generated from Gas Ionisation Chamber Data |
| nea-1867 |
RAPID, RAdial power and burnup Prediction by following fissile Isotope Distribution in the pellet |
| nea-1539 |
RAPRAN, Radionuclide Migration from Waste Glass Release |
| nea-0632 |
RAPVOID, H2O Flow and Steam Flow in Pipe System with Phase Equilibrium |
| nesc0889 |
RAS, Fault Tree Analysis, Reliability, Minimal Cut Sets for Common Cause Failure |
| ccc-0553 |
RASCAL 3.0.5, Radiological Doses from Accidental Release to Atmosphere |
| nesc0758 |
RASE4, Ion Implantation in Solids, Range, Straggling, Energy Deposition, Recoils |
| nea-0475 |
RASPA, Burnup with Fission Products Inventory, Gamma Spectra, Isotopic Power Density |
| csni2300 |
RASPLAV, Refine accident management strategies during a reactor core meltdown |
| ests0050 |
RATAF, Radioactive Liquid Tank Failure |
| ccc-0632 |
RBD, Doses from Radionuclide Inhalation, Ingestion, Wound Uptake from Bioassays |
| nesc1090 |
RCSLK9, PWR Coolant System Leak Rate |
| nea-0168 |
RDMM, Flux Spectra from In-Pile Fast Neutron Activation Experiment |
| ccc-0443 |
REAC*3, Isotope Activation and Transmutation in Fusion Reactors |
| nea-1814 |
REACTORSHIELDING-NMS, Reactor Shielding for Nuclear Engineers by N. M. Schaeffer |
| csni1022 |
REBEKA, Behaviour of a Fuel Bundle Simulator during a Specified Heatup and Flooding Period Results |
| iaea0846 |
REBEL-3, Whole Body and Organ Gamma Doses of Inhomogeneous Phantom by Monte-Carlo |
| ccc-0708 |
REBUS-PC 1.4, Code System for Analysis of Research Reactor Fuel Cycles |
| ccc-0653 |
REBUS3/VARIANT8.0, Code System for Analysis of Fast Reactor Fuel Cycles |
| ests0176 |
RECAP, Replacement Energy Cost for Short-Term Reactor Plant Shut-Down |
| iaea0848 |
RECENT2010, Reconstruction of Cross Sections Data from Resonance Parameters |
| nesc9967 |
RECOG-ORNL, Pattern Recognition Data Analysis |
| nea-0761 |
RECTC/RECTCF, 2nd Order Elliptical Partial Differential Equation, Arbitrary Boundary Conditions |
| nea-0519 |
REDIFFUSION, 1-D Neutron Removal-Diffusion and Gamma Point-Kernel Calculation for Shielding |
| nea-0510 |
REEX-1, U, Pu, Nitric Acid Distribution in Counter Current Pluristage Stripping |
| nesc1065 |
REFCO83, Nuclear Fuel Cycle Cost Economics Using Discounted Cash Flow Analysis |
| nea-0914 |
REFIT, Multilevel Resonance Parameter Least Square Fit of Neutron Transmission, Capture, Fission & Self Indication Data |
| nea-0262 |
REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR |
| nea-1231 |
REFREP, Near-Field Model for Spent Fuel Repository |
| iaea1314 |
RELABEL2010, Labels FORTRAN Statements in ENDF Format Processing Programs |
| nesc0369 |
RELAP-4, Transient 2 Phase Flow Thermohydraulics, LWR LOCA and Reflood |
| nesc0917 |
RELAP-5, Transient 2 Phase Flow Thermohydraulics, LWR LOCA Accidents |
| nea-0437 |
RELAP-UK, Thermohydraulic Transients and Steady-State of LWR |
| nea-0821 |
RELAP/REFLA, Core Reflooding During PWR LOCA |
| nesc0733 |
RELAP3B/MOD110, Flow Temperature Pressure Steam Quality in LWR after LOCA and Accidents |
| nea-0615 |
RELOSS, Reliability of Safety System by Fault Tree Analysis |
| ests0579 |
REMIT, Radiation Exposure Monitoring and Information Transmittal System |
| psr-0482 |
REMIT5.1, Radiation exposure monitoring and information transmittal system |
| nea-0429 |
REMO, Failure Analysis of System with Reparable and Standby Components by Monte-Carlo |
| nea-0101 |
REP-3, Time-Dependent Xe and Sm Poisoning from Space-Dependent Flux Distribution |
| ccc-0586 |
REPRISK-PC, Radioactive Waste Storage Risk Assessment |
| nesc0465 |
RESEND, Infinitely Dilute Point Cross-Sections Calculation from ENDF/B Resonance Parameter |
| nea-0932 |
RESENDD, Resonance Cross-Sections Calculation from ENDF/B-4 and ENDF/B-5 |
| ccc-0786 |
RESRAD 6.5, Residual Radioactive Material Guideline Implementation |
| ests1225 |
RESRAD-BUILD2.36, Residual Radioactive Material Guideline Implementation |
| iaea1286 |
RETRAC, Reactor Core Accident Simulation |
| nea-0979 |
RETRANS, Reactivity Transients in LWR |
| csni1029 |
REWET, PWR LOCA accidents experiments |
| iaea0935 |
REX1-87, MultiGroup Neutron Cross-Sections from ENDF/B |
| iaea0965 |
RGENDF, Conversion of NJOY MultiGroup Cross-Sections to ENDFB-5, EXPANDA, PFCOND, COMPAR Format |
| iaea0969 |
RHEIN, Modular System for Reactor Design Calculation |
| nea-0508 |
RHFPPP, SCF-LCAO-MO Calculation for Closed Shell and Open Shell Organic Molecules |
| ccc-0137 |
RIBD, Fission Products Inventory and Delay Heat in Fast Reactors, with Data Library |
| ccc-0382 |
RIBD-IRT, Isotope Buildup and Isotope Decay from Fission Source |
| nea-0239 |
RIBOT-5, 0-D Burnup for 5 Group BWR or PWR Lattice |
| iaea0929 |
RICANT, Neutron Transport in X-Y Geometry for Isotropic Scattering |
| nesc0453 |
RICE, Energy Exchange Matrix, Damage Cross-Sections, Recoil Energy Spectra from ENDF/B |
| nea-0589 |
RICE-CEGB, Long-Term Actinides and Fission Products Inventory of Irradiated Fuel |
| nesc0720 |
RICE-LASL, Hydrodynamic of Chemically Reactive Mixture by 2-D Navier Stokes Equation |
| nesc9580 |
RICKI, Interactive Gamma Spectra Unfolding with Isotope Identification |
| nea-0234 |
RICM, Resonance Absorption in Multi-Region Slab or Square or Hexagonal Lattice |
| nesc0213 |
RIFF-RAFF, Resonance Integrals in 2 Region Cell, Isotropic Flux and Isotropic Scattering |
| nesc0638 |
RIGEL-4, ENDF/B Utility, Data Retrieval, BCD to BIN Conversion |
| nea-1825 |
RIMACS, Reactor Inspection Main Control System |
| nea-1356 |
RIPP2, H2O Chemistry File Generator for Program PHREEQE |
| ests0185 |
RIPPLE, Incompressible Fluid Dynamics with Free Surfaces |
| ccc-0486 |
RISKAP, Risk Assessment of Radiation Exposure for Population |
| ccc-0623 |
RISKIND, Radiological Risk Assessment for Spent Nuclear Fuel Transportation |
| ccc-0626 |
RIVER-RAD, Radionuclide Transport in Surface Waters |
| nea-1132 |
RKFB, Space-Independent Reactor Kinetics with Temperature Feedback |
| nesc0831 |
RO-75, Reverse Osmosis Plant Design Optimization and Cost Optimization |
| nea-1080 |
RODBURN, Power Profiles and Isotopics in PWR, BWR Fuel Rods |
| 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 |
| ccc-0761 |
RSAC-7.2, Gamma doses, inhalation and ingestion doses, fission products inventory after fission products release |
| nea-0598 |
RSYST, Modular System for Reactor Core and Shielding Problems |
| nesc0245 |
RTS, Non-Equilibrium Reactor Kinetics in Delayed Neutron Regime |
| nea-1835 |
Reactor Shielding Design Manual by Rockwell T. III |
| csni1000 |
S.E.T., CSNI Separate Effects Test Facility Validation Matrix |
| nea-0484 |
S1CALC, Scattering Law for Delta Function or Gaussian Phonon Spectra |
| nea-0402 |
SABINE-3, Neutron Penetration and Gamma Penetration in Reactor Shield for Planar, Spherical, Cylindrical Geometry |
| psr-0242 |
SABRINA, Geometry Plot Program for MCNP |
| nea-1688 |
SACALC2B, Calculates the average solid angle for source-detector geometries |
| nea-1688 |
SACALC_CYL, Calculates the average solid angle subtended by a volume |
| nea-1078 |
SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System |
| ccc-0517 |
SADDE-MOD1, Beta Spectra Evaluation Input Generator for Program VARSKIN |
| nea-0460 |
SAFE-2D/FBM, Elastic Stress Analysis of Mix of Plane and Axial Structure |
| nesc0332 |
SAFE-3D, Stress Analysis of 3-D Composite Structure by Finite Elements Method |
| nesc0251 |
SAFE-AXISYM, Stress Analysis of Axisymmetric Composite Structure by Finite Elements Method |
| nesc0451 |
SAFE-CRACK, Viscoelastic Analysis of Plane and Axisymmetric Concrete System, Finite Elements Method |
| nesc0300 |
SAFE-CREEP, Viscoelastic Analysis of Concrete Structure, Age Temperature and Temperature Dependent Creep |
| nesc0250 |
SAFE-PCRS, Stress Analysis of Axisymmetric Composite Structure by Finite Elements Method |
| nesc0252 |
SAFE-PLANE, Stress Analysis of Planar Structure by Finite Elements Method |
| nesc0253 |
SAFE-SHELL, Stress Analysis of Axisymmetric Thin Shells by Finite Elements Method |
| nesc0674 |
SAFTAC, Monte-Carlo Fault Tree Simulation for System Design Performance and Optimization |
| nea-1779 |
SAGEP-FR, Sensitivity Analysis of Fast Reactor Parameters |
| nea-0212 |
SAHYB-2, Solution of Ordinary Differential Equation with User-Supplied Subroutine |
| nesc0919 |
SALE, Quality Control of Analytical Chemical Measurements |
| nesc0900 |
SALE-2D, 2-D Fluid Flow, Navier Stokes Equation Using Lagrangian or Eulerian Method |
| nesc1069 |
SALE-3D, 3-D Fluid Flow, Navier Stokes Equation Using Lagrangian or Eulerian Method |
| iaea0861 |
SALLY, Dynamic Behaviour of Reactor Cooling Channel by Point Model |
| nesc9849 |
SALT-4, Temperature and Stress from Radioactive Waste Disposal in Bedded Salts |
| ccc-0187 |
SAM-CE, Time-Dependent 3-D Neutron Transport, Gamma Transport in Complex Geometry by Monte-Carlo |
| nesc1120 |
SAMCR, 2-D Elastodynamic Fracture Analysis |
| psr-0158 |
SAMMY, Multilevel R-Matrix Fits to Neutron and Charged-Particle Cross-Section Data Using Bayes' Equations |
| nesc0879 |
SAMPLE, Mean and Standard Deviation and Probability of Given Function by Monte-Carlo |
| nea-0691 |
SAMPO80, Ge(Li) Detector Gamma Spectra Unfolding with Isotope Identification |
| iaea0837 |
SAMSY, Neutron and Gamma Dose Rates and Heat Source for Multilayer Shields |
| nesc9603 |
SANCHO, Quasistatic Large Deformation Inelastic Response of Planar, Axial Solids |
| ccc-0112 |
SAND-2, Neutron Flux Spectra from Multiple Foil Activation Experiment |
| ccc-0361 |
SANDYL, 3-D Time-Dependent and Space-Dependent Gamma Electron Cascade Transport by Monte-Carlo |
| nesc0641 |
SAP-4, Static and Dynamic Linear System Stress Analysis for Various Structures |
| psr-0405 |
SAPHIRE 7.27, Systems Analysis Programs for Hands-On Integrated Reliability Evaluations |
| nea-0520 |
SARAZE-2, Energy Release from Reactivity Transient Fast Reactor Accident |
| nea-0204 |
SASSI, Total and Differential Elastic and Inelastic Neutron Cross-Sections by Hauser-Feshbach |
| nea-1694 |
SATIF/CYCLO-RADSAFE, Health Physics and Radiological Safety of Cyclotrons 10-250 MeV |
| iaea0917 |
SC2N3N, (n-2n) and (n-3n) Cross-Sections Systematics |
| ccc-0785 |
SCALE 6.1, Modular system for criticality, shielding, source term, fuel depletion/decay, inventories, reactor physics |
| nea-1405 |
SCALPLO, Plotting of Flux Output from SCALE Program |
| psr-0352 |
SCAMPI, Problem Dependent Library Preprocessing in AMPX Format |
| nesc1119 |
SCANS, Shipping Cask Design Safety Analysis |
| ccc-0418 |
SCAP-82, Single Scattering, Albedo Scattering, Point-Kernel Analysis in Complex Geometry |
| nea-0444 |
SCARF-4, Nonlinear Stresses in Pressure Vessel Liner with Plastic Behaviour Simulation |
| nea-0829 |
SCAT-2, Cross Sections and Angular Distributions for Spherical Nuclei by Optical Model |
| nea-0829 |
SCAT-2B, Spherical, Optical Model Cross Sections Calculation for N, P, D, T, He3, He4, Heavy Ions |
| iaea0913 |
SCENARIOS, Simulation of Reactor Introduction and Operation Scenario Needs |
| nea-0431 |
SCEPTIC, Pressure Drop, Flow Rate, Heat Transfer, Temperature in Reactor Heat Exchanger |
| nesc0802 |
SCHAFF, Single-Phase Flow, Heat Transfer in Porous Media, Geothermal Energy System |
| nea-0994 |
SCINFI, Quenching Function of Beta-Ray Liquid Scintillation Detectors |
| psr-0267 |
SCINFUL, Scintillation Neutron Detector Response by Monte-Carlo |
| nea-1755 |
SCIP, Radioactive Surface Contamination Investigation Program |
| psr-0210 |
SCOPE, Shipping Cask Optimization and Parametric Evaluation |
| nea-0235 |
SCOPERS-2, BWR and PWR Core Performance Simulation |
| nea-0498 |
SCORCH-B2, BWR Core Heating During LOCA |
| nea-0407 |
SCORE-4, 2-D Removal Diffusion in X-Y or R-Z Geometry for Rectangular Shields |
| ests0015 |
SCORE-EVET, 3-D Hydraulic Reactor Core Analysis |
| nea-0537 |
SCOTCH, 1-D 2 Group HTGR Core Kinetics with Temperature Transients and Fluid Dynamic |
| nesc1002 |
SCREEN, Statistical Sensitivity Ranking of Program Input Variables |
| nea-1540 |
SCRELA, LOCA Analysis of Super-Critical Light-Water Reactors |
| nea-0865 |
SCRIMP, Steady-State Thermohydraulics of HTGR Subchannel |
| nesc9717 |
SCWEB, Scientific Workstation Evaluation Benchmark |
| ccc-0620 |
SEECAL-2.0, Specific Effective Energy in Human Body Due to Radiation |
| nesc1063 |
SEISIM-1, Seismic Probabilistic Risk Assessment |
| nea-0654 |
SELFS-3, Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-2 |
| csni0027 |
SEMISCALE/S-06-3(LV), Thermal and Hydraulic Phenomena with LOCA in U-Tube PWR |
| csni0028 |
SEMISCALE/S-06-3(SV), Thermal and Hydraulic Phenomena with LOCA in U-Tube PWR |
| csni0023 |
SEMISCALE/S-IB-3, Semiscale MOD-2A, 21.7% Communication Cold Leg Break LOCA Experiment |
| csni0024 |
SEMISCALE/S-PL-3E-LV, Scaled Reference 4-Loop PWR Experiment, Loss of Offsite Power Simulation |
| csni0025 |
SEMISCALE/S-PL-3E-SV, Scaled Reference 4-Loop PWR Experiment, Loss of Offsite Power Simulation |
| csni0077 |
SEMISCALE/TEST1011, Large break LOCA blowdown starting from isothermal conditions of the primary coolant loop |
| csni0026 |
SEMISCALE/UT-1, Semiscale MOD-2A, 10% Communication Cold Leg Break LOCA Experiment |
| nesc9438 |
SENSIT MUSIG COMSEN, Sensitivity Test Analysis |
| ccc-0405 |
SENSIT, Integral Response Sensitivity from Neutron Cross-Sections, Gamma Cross-Sections Errors |
| ccc-0729 |
SERA-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 |
| iaea0925 |
SHARDA, Thermal Reactor Isotope Irradiation Analysis |
| ests0204 |
SHC, Seismic Hazard Assessment for Eastern US |
| nesc0452 |
SHELL-5, Elastic Stress Analysis of 3-D Thin Shells Using Finite Elements Method |
| iaea1287 |
SHIELD, Monte-Carlo Code for Simulating Interaction of High Energy Hadrons with Complex Macroscopic Targets |
| iaea1391 |
SHIELDER, Gamma shielding calculations of radionuclides emitting photons 0-5 to 10 MeV |
| ccc-0379 |
SHIELDOSE, Doses from Electron and Proton Irradiation in Space Vehicle Al Shields |
| nesc0795 |
SHOCK, Nonlinear Dynamic Structure Analysis, Spring and Mass Model, Runge-Kutta-Gill Method |
| nea-0538 |
SHOSPA-MOD, Hot Spot Factors for Fuel and Clad, Hot Channel Factors |
| iaea0826 |
SHOVAV, Space-Dependent and Time-Dependent Neutron Diffusion with Temperature Feedback in Slab Geometry |
| nea-0466 |
SHREDI, Neutron Flux and Neutron Activation in 2-D Shields by Removal Diffusion |
| nea-0852 |
SICOS, 2-D Time-Dependent Creep Calculation of Plane or Axisymmetric Concrete Structure |
| iaea1416 |
SIGACE, Code for Doppler broadening of ACE-formatted files |
| ccc-0118 |
SIGMA/B, Doses in Space Vehicle for Multiple Trajectories, Various Radiation Source |
| iaea0854 |
SIGMA1-2010, Doppler Broadening ENDF Format Linear-Linear. Interpolated Point Cross Section |
| nea-0571 |
SIGMARZ, Stress Analysis of Axisymmetric or Plane Structures |
| nesc1082 |
SIGPI, Probabilistic System Performance by Fault-Tree Analysis |
| ests0238 |
SIMION, Electrostatic Lens Analysis and Design |
| nesc9593 |
SIMPLE, 2-D Hydrodynamic, Heat Flow Benchmark |
| nea-0319 |
SIMPLED-4, 1-D Neutron Diffusion in Spherical, Cylindrical, Planar Geometry from JAERI Fast-Set and ABBN |
| ests0767 |
SIMSOL, Multiphase Fluid and Heat Flow in Porous Media |
| nea-1552 |
SINBAD ACCELERATOR, Shielding Benchmark Experiments |
| nea-1553 |
SINBAD FUSION, Neutronics Benchmark Experiments |
| nea-1517 |
SINBAD REACTOR, Shielding Benchmark Experiments. |
| psr-0139 |
SIOB, Least Square Fit of Neutron Transmission Data Using Multilevel Breit-Wigner |
| nesc0687 |
SITE-2, Power Plant Siting, Cost, Environment, Seismic and Meteorological Effects |
| nea-1570 |
SITE-94, Biosphere Model for SKI Project on the island of Aspro |
| nea-0770 |
SITO, Environmental Impact of Major Industrial Activities |
| iaea1283 |
SIXPAK2010, ENDF Format Double Differential Cross Section Converter to Single Differential Format |
| nea-0905 |
SIXTUS-2, 2-D MultiGroup Diffusion in Hexagonal Geometry with Intranodal Solution |
| nea-1426 |
SIXTUS-3, 3-D Nodal Neutron Diffusion Criticality in Hexagonal Geometry |
| nea-1577 |
SKETCH-N 1.0, Solve Neutron Diffusion Equations of Steady-State and Kinetics Problems |
| ccc-0289 |
SKYSHINE, Dose Rate Outside Concrete Steel Building from 6 MeV Gamma by Monte-Carlo |
| ccc-0646 |
SKYSHINE-KSU, Gamma Skyshine Doses by Integral Line-Beam Method |
| nesc0581 |
SLADE-D, Transient Dynamic Response of Elastic Shells by Finite Elements Method |
| nesc9776 |
SLAP, Large Sparse Linear System Solution Package |
| nea-1081 |
SLAROM, Neutron Flux Distribution and Spectra in Lattice Cell |
| ests0181 |
SLATEC-4.1, Subroutine Library for Solution of Mathematical Problems |
| nesc9770 |
SLIB77, Source Library Data Compression and File Maintenance System |
| ccc-0704 |
SLIDERULE 1.0,Slide Rule for direct radiation exposure approximation in criticality accidents |
| nesc1077 |
SMACS, Probabilistic Seismic Analysis Chain with Statistics |
| nea-1767 |
SMAFS, Steady-state analysis Model for Advanced Fuelcycle Schemes |
| nea-1046 |
SMART, Radiation Dose Rates on Cask Surface |
| ccc-0602 |
SMART-BNL, Offsite Radionuclide Air Concentration from Reactor Accident |
| csni1017 |
SMD/12R305C, Steady state critical flow in nozzles, medium to high pressure conditions |
| nea-0026 |
SMOG, Optical Model Neutron Cross-Sections with Fox-Goodwin Integral Method |
| nea-0430 |
SNAP, MultiGroup 3-D Neutron Diffusion in X-Z, R-Theta-Z, Hexagonal-Z, Triangular-Z Geometry |
| nesc0189 |
SNC, Sn Constant Calculation for Program DSN and TDC |
| psr-0345 |
SNL-SAND-II, Neutron Flux Spectra from Multiple Foil Activation Analysis |
| nesc0521 |
SOCOOL-2, Molten Materials Na Coolant Interaction, Temperature and Pressure Transient |
| nesc0764 |
SOERP, Statistics and 2nd Order Error Propagation for Function of Random Variables |
| nesc0559 |
SOFIRE-2, Containment Temperature and Pressure During Na Pool Fire, 1-Cell or 2-Cell Analysis |
| nesc0832 |
SOLA-DF, Time-Dependent 2-D 2 Phase Flow, Eulerian Method with Various Boundary Conditions |
| nesc0723 |
SOLA-ICE, Compressible Fluid Flow Transients, 2-D Planar, Cylindrical Geometry, Eulerian Method |
| nesc0859 |
SOLA-LOOP, Transient 2 Phase Flow in Networks of 1-D Components |
| nesc0651 |
SOLA-SURF, 2-D Plane, Axisymmetric, Incompressible Flow Navier Stokes Equation for Transient |
| nesc0948 |
SOLA-VOF, 2-D Transient Hydrodynamic Using Fractional Volume of Fluid Method |
| nesc9944 |
SOLGASMIX-PV, Chemical System Equilibrium of Gaseous and Condensed Phase Mixtures |
| nea-1826 |
SOLTRAN, solving multi-dimensional simplified P2 transport and diffusion problems of hexagonal geometry in fast reactors |
| uscd1100 |
SOLUPLOT, Eh-pH Diagram, a02-pH Diagram Plots for Aqueous Chemical Systems |
| nesc0662 |
SOLVEX, Dynamic and Steady-State Mixer-Settler and Centrifugal Contactor Behaviour |
| nea-1641 |
SONATINA, Predicts Behaviour of Prismatic HTGR Core under Seismic Excitation |
| psr-0174 |
SORA, Radionuclide Analysis Data Storage and Retrieval |
| nea-0187 |
SOREX-1, Worst Accident Simulation in Sora Pulsed Fast Reactor |
| nea-0450 |
SOTHIS, PWR Fuel Cycle Equilibrium Cost Evaluation |
| ccc-0661 |
SOURCES-4C, Calculating Alpha, N, Fission, Delayed Neutron Sources and Spectra |
| ccc-0120 |
SPACETRAN, Radiation Leakage from Cylinder with ANISN Flux Calculation |
| iaea0895 |
SPAGAF, PWR Fuel, Cladding Behaviour with Fission Products Gas Release |
| nea-0219 |
SPANDE, Stress Analysis of General Spaceframe and Pipework |
| ccc-0228 |
SPAR, High-Energy Muon, Pion, Heavy Ion Stopping-Powers and Ranges |
| ccc-0148 |
SPARES, Program System for Space Radiation Environment and Shielding System Evaluation |
| nea-0468 |
SPARK, Time-Dependent 1-D, 2-D, 3-D Diffusion with Heat Transfer and Feedback |
| nea-0219 |
SPATAM, Tilt Angle Calculation of Framework for Program SPANDE |
| iaea1332 |
SPEC, Neutron and Charged-Particle Reactions by Optical Model, Evaporation Model |
| nesc9641 |
SPECFUN1, Portable Special FORTRAN Routines with Test Drivers |
| psr-0263 |
SPECTER-ANL, Neutron Damage for Material Irradiation |
| iaea1433 |
SPECTRA2010, Convert model and general tabulation to linearized spectra (MF=5) |
| nea-1165 |
SPEEDI,EXPRESS, Radiation Dose from Plume Release in Nuclear Accident |
| nea-0374 |
SPES, Fuel Cycle Optimization for LWR |
| csni0075 |
SPES/SP-FW-02, Total Loss Feedwater with Emergency Feed Water Delayed in SPES facility |
| nea-0548 |
SPIRIT, Plot of Geometry and Results of 2-D Finite Elements Calculation |
| ests0054 |
SPIRT, Stress Strains from Transient Pressure |
| nea-0462 |
SPLINE, Spline Interpolation Function |
| nea-0609 |
SPLOSH-3, 1-D Time-Dependent Coupled Neutron Kinetics Thermohydraulics for PWR Transient |
| nesc9736 |
SPLPKG WFCMPR WFAPPX, Wilson-Fowler Spline Generator for Computer Aided Design And Manufacturing (CAD/CAM) Systems |
| nea-0157 |
SPM-046, Reactor Kinetics by 1 Group Diffusion Calculation in R-Z Geometry |
| nea-1173 |
SPOP-4, Uncertainty and Sensitivity Analysis Monte-Carlo Program with Input from PREP |
| ccc-0460 |
SPOT1, Gamma-Ray Dose Rate from Cylindrical Source Volume |
| nesc0279 |
SPOTS, Library Generator for Program LEOPARD from Cross-Sections Data |
| nesc0716 |
SPRAY-3, Thermodynamics and Heat Transfer of Na Sprays in LMFBR after Pipe Failure |
| psr-0266 |
SPUNIT, Multisphere Neutron Spectra Unfolding |
| nea-0414 |
SQUID-360, 2-D Neutron Diffusion in X-Y and R-Z Geometry with Criticality Search and Constant Neutron Source |
| psr-0533 |
SQUIRT 1.1, predicts leakage rate and crack area for cracked pipes in nuclear power plants |
| ests1052 |
SQUIRT, Seepage in Reactor Tube Cracks |
| nea-0842 |
SRAC-95, Cell Calculation with Burnup, Fuel Management for Thermal Reactors |
| nea-0919 |
SRIM-2008, Stopping Power and Range of Ions in Matter |
| iaea1382 |
SRNA-2K5, Proton Transport Using 3-D by Monte Carlo Techniques |
| nea-0684 |
SSYST, Modular System for Transient Fuel Rod Behaviour Under Accident Condition |
| nesc9850 |
STAFAN, Fluid Flow, Mechanical Stress in Fractured Rock of Nuclear Waste Repository |
| nea-1725 |
STAMPI, Application to the Coupling of Atmosphere Model (MM5) and Land-surface Model (SOLVEG) |
| uscd1218 |
STANEF, ENDF/B Book-keeping Operations for ENDF Format Files |
| iaea0971 |
STAPRE-H95, Evaporation and Pre-Equilibrium Model Reaction Cross-Sections Calculations |
| nea-0461 |
STAPREF, Nuclear Reactions Cross-Sections by Evaporation Model, Gamma-Cascades |
| iaea0882 |
STAR, Fuel Management of BWR |
| psr-0330 |
STARCODES, Stopping Power and Ranges for Electrons, Protons, He |
| nea-0986 |
STATCAT, Statistical Analysis of Parametric and Non-Parametric Data |
| nea-0908 |
STATISTICS, Program System for Statistical Analysis of Experimental Data |
| nesc9749 |
STATLIB, Interactive Statistics Program Library of Tutorial System |
| nea-0352 |
STAX-2, Neutron Scattering Cross-Sections by Optical Model and Moldauer Theory with Hauser-Feshbach |
| psr-0113 |
STAY-SL, Dosimetry Unfolding with Activation, Dosimetry, Flux Error Calculation |
| nea-0055 |
STDY-3, Steady-State Parallel Channel Thermal Analysis of PWR |
| nea-0703 |
STEADY-ACE, 3-D Neutronics and Multichannel Thermohydraulics Analysis of BWR |
| nesc0487 |
STEAM-67, Thermodynamics Properties of H2O and Steam from ASME Tables (1967) |
| nesc0749 |
STEEP-4, Fusion Reaction Rates for Maxwellian and Slowing-Down Plasma Ion Distribution |
| nea-0575 |
STESTA, Steady-State State-Variable Profiles of Thermohydraulic Piping System |
| csni2007 |
STEX-II, International Steam Explosion Experimental Data Base |
| nesc9852 |
STFLO, Steady-State H2O Flow in Porous Media |
| nesc0652 |
STFODE-COLODE, 1st Order Stiff Ordinary Differential Equation System by Collocation Method |
| nea-0549 |
STIGMA STIG STEGT STAGT STABA, Stress Analysis of Dragon HTR Graphite Structure |
| iaea0900 |
STOFFEL-1, Steady-State In-Pile Behaviour of Cylindrical H2O Cooled Oxide Fuel Rod |
| iaea0970 |
STOPOW, Stopping Power of Fast Ions in Matter |
| ccc-0067 |
STORM, Radiation Hazard of Solar Flares for Space Vehicles |
| nea-0993 |
STRADE, Stratified Random Design for Reactor Safety Analysis |
| nesc0539 |
STRAP-2, Stress Analysis of Structure with Static Loading by Finite Elements Method |
| nesc0539 |
STRAP-D, Stress Analysis of Structure with Time-Dependent Loading by Finite Elements Method |
| nea-0349 |
STRESSPLOT, CALCOMP Plot of 2-D Finite Elements Calculation |
| iaea0943 |
STRIMP, Impurity Evolution in Tokamak Fusion Reactor Discharge |
| nea-0253 |
STYLE, Steam Cycle Heat Balance for Turbine Blade Design in Marine Operation |
| nesc0924 |
SUBDOSA, External Gamma, Beta Doses from Radionuclide Release into Atmosphere |
| iaea1176 |
SULSA, New Method for Neutron Spectrum Unfolding Problem |
| nesc0056 |
SUMMIT, Energy Transfer Diffusion Cross-Sections, Crystalline Moderator, Phonon Expansion |
| nesc0638 |
SUMUP-4, ENDF/B Utility, Partial Cross-Sections Sum Check Against Tot Cross-Sections |
| psr-0282 |
SUPERDAN-PC, Dancoff Factor for Spherical, Cylindrical, Slab Geometry |
| psr-0013 |
SUPERTOG, MultiGroup Cross-Sections Generator from ENDF/B for Programs GAM, ANISN, DOT |
| iaea0894 |
SUPERTOG-LTT, SUPERTOG with Tabular Elastic Scattering Anisotropy from ENDL |
| nesc9608 |
SUPES, Engineering Sciences Utilities Program Library |
| nesc0731 |
SUPORT, Solution of Linear 2 Point Boundary Value Problems, Runge-Kutta-Fehlberg Method |
| nesc0853 |
SURGTANK, Steam Pressure, Saturation Temperature or Reactor Surge Tank |
| nea-1151 |
SUSD, Sensitivity and Uncertainty in Neutron Transport and Detector Response |
| nea-1628 |
SUSD3D, 1-, 2-, 3-Dimensional Cross Section Sensitivity and Uncertainty Code |
| ccc-0248 |
SWAN-PPL, Fusion Reactor 1-D Particle Transport Optimization |
| ccc-0204 |
SWANLAKE, Cross-Sections Sensitivity Analysis for 1-D Discrete Ordinate Calculation |
| nesc0828 |
SWAP-9, 1-D Stress Analysis for Hydrostatic and Elastic Plastic Materials |
| nea-1698 |
SWAT, Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2 |
| nesc9811 |
SWENT, 3-D Fluid, Heat, Radionuclide Transport in Heterogeneous Geologic Medium |
| nesc0973 |
SWIFT, 3-D Fluid Flow, Heat Transfer, Decay Chain Transport in Geological Media |
| ests0682 |
SWIMS, Sigmund and Winterbon Multiple Scattering of Ion Beams |
| nesc0713 |
SYN-3D, 2-D and 3-D Neutron Diffusion Static Eigenvalues, Single Channel Spatial Flux Synthesis |
| nea-0594 |
SYNTH-C, Steady-State and Time-Dependent 3-D Neutron Diffusion with Thermohydraulic Feedback |
| iaea1383 |
SYRCO-1, 1-D, 2 Groups Multi-Zone Interactive Diffusion Code |
| nea-1023 |
SYVAC, Risk Assessment from Underground Radioactive Waste Disposal in UK |
| nesc9766 |
T-HEMP3D, 3-D Time-Dependent Elastic Plastic Flow |
| ests0219 |
T2VOC, H2O, Air, VOC Flow Simulation in Porous Multidimensional Media |
| nesc0408 |
TAC-2D, Steady-State and Transient Heat Transfer in X-Y, R-Z or R-Theta Geometry |
| nesc0414 |
TAC-3D, 3-D Steady-State and Transient Heat Transfer in X-Y-Z and R-Theta-Z Geometry |
| nesc9838 |
TAC0-3D, 3-D Linear or Nonlinear, Steady-State or Transient Heat Transfer |
| iaea0872 |
TACHY, BWR Fuel Management by 2-D Coarse Mesh Neutron Diffusion |
| nesc1113 |
TACT-5, Doses of Radioactivity Release from Reactor Core into Environment |
| nea-0532 |
TAFE, 2-D Steady-State Heat Conduction for Structure with Gas Gaps |
| nea-0531 |
TAFEST, 2-D Transient Heat Conduction |
| nea-1737 |
TALYS-1.2, computes nuclear reactions cross-sections, yields and spectra via a comprehensive set of nuclear models |
| psr-0308 |
TAM3, Monte-Carlo Sensitivity and Uncertainty Analysis of Radium in Lake Contamination Model |
| nesc9566 |
TAP-LOOP, Steady-State and Transient Thermal Analysis of Closed Test Loops |
| nea-1301 |
TAPE, General Copy Utility for VAX/VMS and IBM Tapes |
| nea-0556 |
TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements |
| ccc-0638 |
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 |
| iaea1272 |
THACT-RR, Analysis of Thermal Hydraulics Transients in Research Reactor Core |
| nea-0774 |
THALES, Thermohydraulic LOCA Analysis of BWR and PWR |
| nea-1098 |
THARC-S, Rod Bundle Thermohydraulic Transients of LMFBR for Single Phase Conditions |
| nea-0634 |
THERLIB, Library Generated for THERMOS from FACEL Library |
| nesc9940 |
THERMIT, 3-D Thermo-Hydraulics of BWR and PWR |
| nea-0634 |
THERMLIB, Generator and Edit of Program THERMOS-OTA Library |
| 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 |
| nesc0184 |
THERMOS-ANL, 1-D Integral Transport for Neutron Spectra, Slab and Cylinder |
| nea-0628 |
THERMOS-OTA, Thermal Flux by Integral Transport |
| nea-0411 |
THESEE-3, Orgel Reactor Performance and Statistic Hot Channel Factors |
| nesc0512 |
THETA-1B, Fuel Rod Temperature Distribution by 2-D Diffusion, Heat Transfer to Coolant, LWR LOCA |
| csni1016 |
THETIS, Single Phase Cooling, Forced and Gravity Reflood, Level Swell Experiments |
| nea-0869 |
THIDA, Transmutation, Hazard Potential, Dose Rate in Fusion Reactor |
| nea-0869 |
THIDA-2, Transmutation, Activation, Decay Heat, Dose Rate in Fusion Reactor |
| nea-0377 |
THREAT, 3-D Steady-State or Transient Heat Diffusion in Multi-Region Prism |
| nesc0504 |
THRES-2, Nuclear Induced Particle Emission Cross-Sections from Statistical Models |
| nea-0658 |
THRUSH, Thermal Neutron Coherent and Incoherent Scattering Kernels by Phonon Expansion |
| nea-0997 |
THT, 3-D Coarse Mesh LWR Bundle Fluxes and Power with Discontinuity Factors |
| nea-0778 |
THYDE-B2, Thermohydraulic Transients During LOCA of BWR |
| nea-0779 |
THYDE-P, PWR LOCA Thermohydraulic Transient Analysis |
| nea-1592 |
TIBSO, Nuclear Transitions and Radioactivity Migration in Technological System |
| ests0643 |
TIDY6.21, Reformatting of FORTRAN Source Programs |
| nea-1077 |
TIME-2, Radioactive Waste Disposal Climatic Change Risk Assessment |
| nesc0756 |
TIMEX, 1-D Time-Dependent MultiGroup Transport Theory with Delayed Neutron, Planar Cylindrical and Spherical Geometry |
| nea-0387 |
TIMOC-72, 3-D Time-Dependent Homogeneous or Inhomogeneous Neutron Transport by Monte-Carlo |
| nea-0619 |
TIMOC-ESP, Time-Dependent Response Function by Monte-Carlo with Interface to Program TIMOC-72 |
| nea-0804 |
TIMS-1, MultiGroup Cross-Sections of Heavy Isotope Mixture with Resonance from ENDF/B |
| nea-0701 |
TIRION-4, Atmospheric Dispersion of Radioactive Materials for Various Weather Conditions |
| ccc-0759 |
TITAN 1.24, A Three-Dimensional Deterministic Radiation Transport Code System |
| csni0029 |
TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA |
| csni0030 |
TLTA/6432, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA |
| ests0551 |
TMAP4, Tritium Migration Analysis Program Version 4 |
| csni2012 |
TMI-VIP, Three Mile Island Reactor Pressure Vessel investigation OECD/NEA Project |
| ests0219 |
TMVOCV1.0, Multicomponent, multiphase, nonisothermal flows of water, soil gas, volatile organic chemicals (VOCs) |
| psr-0298 |
TNG1, Multistep Statistical Model Hauser-Feshbach |
| nesc9863 |
TOEPLITZ, Solution of Linear Equation System with Toeplitz or Circulant Matrix |
| nesc0561 |
TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor |
| nesc0561 |
TOKMINA-2, Total Power for Tokamak Fusion Reactor |
| nesc0627 |
TOODY-2, Lagrangian Nonlinear Wave Propagation in 2-D X-Y or Cylindrical Geometry |
| nesc1056 |
TOOLPACK1, Tools for Development and Maintenance of FORTRAN 77 Program |
| nesc1019 |
TOP-DRAWER, Histograms, Scatterplots, Curve-Smoothing |
| nesc9801 |
TOPAZ, 2-D Plane or Axisymmetric Heat Conduction Analysis |
| nesc9599 |
TOPAZ-3D, 3-D Steady-State or Transient Heat Transfer by Finite Element Method |
| nesc9669 |
TOPAZ-SNLL, Transient 1-D Pipe Flow Analysis |
| nesc9801 |
TOPAZ2D, 2-D Finite Element Method Heat-Transfer and Electrostatic and Magnetostatic (E&M) Potential Field Program |
| iaea0909 |
TOPIC-RUM, Plasma Impurities in Tokamak Reactor by MHD Method |
| nea-1406 |
TOPICS-B, Neutron and Gamma Cross-Sections Library Handling in FIDO Format |
| nesc0599 |
TOPLYR-2, Open Channel H2O Flow Temperature, Distant Source, Time-Dependent Boundary Conditions |
| nesc1093 |
TORAC, Flows, Pressure, Materials Transport within Structure During Tornado |
| ccc-0543 |
TORT, 2-D 3-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration |
| nea-0486 |
TOTEM, Demand Assessment for Nuclear Power Plants and Conventional Power Plants |
| nesc1098 |
TOUGH, Unsaturated Groundwater Transport and Heat Transport Simulation |
| ests0219 |
TOUGH2, Unsaturated Ground Water and Heat Transfer |
| ests0219 |
TOUGHREACTV1.2, Chemically reactive non-isothermal flows of multiphase fluids in porous and fractured media |
| nesc9710 |
TOXRISK, Toxic Gas Release Accident Analysis |
| nea-1024 |
TP2, Calculation of Reactivity and Kinetic Parameter by 2-D Neutron Transport. Perturbation Theory |
| nea-0900 |
TPHEX, MultiGroup Neutron Flux in Homogeneous Hexagonal LWR Cells |
| nea-1070 |
TPLOT, Interactive Postprocessor of Transient Structure Problems |
| nea-1155 |
TPTRIA, Reactivity for 2-D Triangular Geometry by Transport Perturbation Theory |
| nesc0836 |
TRAC, Thermohydraulics, Reactor Kinetics, 2 Phase Flow LOCA Analysis |
| nesc1031 |
TRAC-BD1, LOCA Analysis of BWR with 3-D Pressure Vessel and Multi Bundle Fuel Model |
| nea-1593 |
TRAC-PF1/EN MOD 3, Best Estimate Coupled 3-D Neutronics-Thermalhydraulics |
| nea-1291 |
TRANS-ACE, Radioactive Materials Transport in Reprocessing Plant Fire Accident |
| nesc0268 |
TRANS-FUGUE-1, Single Channel 2 Phase Flow Heat Transfer after Boiling |
| nea-0953 |
TRANSHEX, 2-D Thermal Neutron Flux Distribution from EpiThermal Flux in Hexagonal Geometry |
| nesc0791 |
TRANSPORT, Charged Particle Beam Transport 1st Order and 2nd Order Optical Analysis |
| iaea1209 |
TRANSV2, LOCA and Steady-State Thermohydraulic Analysis of MTR |
| psr-0317 |
TRANSX-2.15, Neutron Gamma Particle Transport Tables from MATXS Format Cross-Sections |
| nea-0745 |
TRAPSCO-2, Pressure and Temperature Transients in PWR Subcompartments During LOCA |
| nea-0807 |
TRAWA, LWR Dynamic by Coupled Neutron Diffusion and Thermohydraulics Calculation |
| nea-0117 |
TRAWS-4, Axial Flux Distribution for Control Rod Variations |
| iaea0942 |
TRAX, Resolution Matrix of Slow Neutron Spectrometers |
| nea-0668 |
TRD-3, In-Core and Out-Core Neutron Flux, Gamma Flux by 2-D Removal Diffusion in Cylindrical Geometry |
| nesc1021 |
TREDRA, Minimal Cut Sets Fault Tree Plot Program |
| iaea0833 |
TREEDE, Point Fluxes and Currents Based on Track Rotation Estimator by Monte-Carlo Method |
| nea-0361 |
TRESS, Triangular Mesh Stress and Strain in R-Z, X-Y Geometry for Various Load and Temperature |
| ccc-0293 |
TRIDENT, 2-D Neutron Transport for Homogeneous and Inhomogeneous Problems in X-Z, R-Z Geometry, Anisotropic Scattering |
| iaea0804 |
TRIFIDO, Decay Constant and Prompt Neutron Calculation from Pulsed Neutron Experiment |
| iaea1214 |
TRIGAC, Flux and Power Distribution and Burnup for TRIGA Reactor |
| iaea1370 |
TRIGLAV, Research Reactor Calculations |
| nea-0384 |
TRIGON, 2-D Homogeneous and Inhomogeneous Fixed Source Neutron Diffusion for Triangular or Hexagonal Mesh |
| nesc1028 |
TRIPM, Isothermal Transport and Decay of Radionuclides in Aquifer |
| nea-1716 |
TRIPOLI-4.3.3 & 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 |
| nea-0233 |
TURBINA, Reheat Steam Turbine Generator Design with Preheater and Condenser |
| nea-0581 |
TURBPLANT, 1-D Steady-State Model of Power Reactor Steam Turbine Components |
| nesc0042 |
TUZ, Resonance Integrals in Unresolved Region, Various Temperature, From Porter-Thomas Distribution |
| nea-0471 |
TVEDIM, 2-D Homogeneous and Inhomogeneous Neutron Diffusion for X-Y, R-Z, R-Theta Geometry |
| nesc0809 |
TVENT, 1-D Incompressible Flow for Pressure Transients in Ventilation System |
| ccc-0547 |
TWODANT-SYS, DANTSYS3.0, 1-D, 2-D, 3-D MultiGroup Discrete Ordinate Method Transport |
| nesc0712 |
TWODEE-2/MOD3, 2-D Time-Dependent Fuel Elements Thermal Analysis after PWR LOCA |
| nesc0358 |
TWOTRAN-2, 2-D MultiGroup Transport in X-Y, R-Z, R-Theta Geometry with Anisotropic Scattering |
| ccc-0195 |
TWOTRAN-GG-FC, General Geometry 2-D Transport with 1st Collision Source Calculation |
| ccc-0195 |
TWOTRAN-GG-VW, General Geometry 2-D Transport, Variable-Weight Diamond Difference |
| nea-1682 |
U3-U5-PU9-CRITICALS, Critical Dimensions of Systems containing U235, Pu239, and U233 |
| nesc9668 |
UCBNE, Radionuclide Migration in Porous Media |
| nesc9667 |
UCBNE25, Radionuclide Migration in Geologic Media |
| nesc0824 |
UDAD, Radiation Exposure to Man at Uranium Processing Plant |
| ests0404 |
UHS, Ultimate Heat Sink Cooling Pond Analysis |
| psr-0015 |
UKE, Format Conversion from UKNDL to ENDF/B |
| nea-1665 |
UMG 3.3, Analysis of data measured with spectrometers using unfolding techniques |
| nea-1139 |
UNC32/33, Covariance Matrices from ENDF/B-5 Resonance Parameter Uncertainties |
| nea-0175 |
UNCLE, Crystal Scattering Kernel with Coherent Scattering by Butler Approximation |
| iaea1242 |
UNF, Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials |
| iaea1177 |
UNIFY, Fast Neutron Cross-Sections and Spectrum for Structural Materials |
| ests0827 |
UNSPEC, X-Ray Spectrum Unfolding |
| iaea0959 |
UPEAK, General Experimental Spectra Analysis Program |
| psr-0245 |
UPEML, Computer Independent Emulator of CDC Update Utility |
| csni1007 |
UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA |
| csni1004 |
UPTF/TEST5A/RUN063, Steam/Water Flow Phenom.Blowdown PWR Cold Leg Break LOCA |
| csni1005 |
UPTF/TEST8A/RUN112, Flow Patterns in Hot or Cold Leg, PWR Large Break LOCA |
| csni1006 |
UPTF/TEST8B/RUN111, Flow Patterns in Hot or Cold Leg, PWR Large Break LOCA |
| psr-0281 |
URR, Cross-Sections, Selfshielding for Fertile and Fissile Isotopes in Unresolved Region |
| ests0333 |
USINT, High Temperature Heat and Mass Transfer on Concrete Surfaces in LMFBR |
| nesc9848 |
UTAH-2, Thermoplastic Response in Anisotropic Rock |
| uscd1150 |
UTAP, U Tailings Assessment Program |
| ccc-0500 |
UTMTOX, Toxic Chemical Transport in Atmosphere, Ground Water, Sediments |
| nea-0356 |
UTOE, UKNDL to ENDF/B Format Conversion with Log-Log Interpolation and Angular Distribution Tables |
| nea-0587 |
UTSG, Steady-State and Transients of Vertical U-Tube Steam Generator |
| ccc-0613 |
VALE-1.1, 2-D, 3-D MultiGroup Neutron Diffusion for Triangular Problems |
| nesc0264 |
VARI-QUIR-3, 2-D MultiGroup Steady-State Neutron Diffusion in X-Y R-Z or R-Theta Geometry |
| nesc0755 |
VARR2 VARRLXSG, 2-D Transient Fluid Flow and Heat Transfer in X-Y and Cylindrical Geometry |
| ccc-0522 |
VARSKIN 3 V3.1.0, Dose Calculation for Skin Contamination, with Sadde Input Generator |
| ccc-0781 |
VARSKIN 4 V4.0.0, Dose Calculation for Skin Contamination, with Sadde Input Generator |
| ests0752 |
VCODE, Ordinary Differential Equation Solver for Stiff and Non-Stiff Problems |
| ccc-0262 |
VCS, Radiation Protection Factors in Vehicles by Monte-Carlo |
| uscd1239 |
VENTEASY, Criticality search for a desired Keffective by adjusting dimensions, nuclide concentrations, or buckling |
| ccc-0654 |
VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup |
| nesc0511 |
VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions |
| nesc9826 |
VERTPAK-1, Fluid Flow, Rock Deformation, Solute Transport in Porous Media |
| nea-1856 |
VESTA 2.0.3, Monte Carlo depletion interface code |
| psr-0311 |
VIDEO-PC, SVGA Routines for FORTRAN on PC |
| ccc-0754 |
VIM 5.1, Steady-State 3-D Neutron Transport Using ENDF/B or Multigroup Cross Sections |
| nesc0510 |
VIM, 3-D Monte-Carlo Analysis of Fast Critical Assemblies Using Point Cross-Sections |
| uscd1240 |
VIM_NC, VIM color syntax for Nuclear Codes: NJOY, DRAGON, PARTISN, TORT, MONK, and MCNP |
| iaea0932 |
VIRGIN2010, Calculates Uncollided Neutron Flux and Neutron Reactions from Transmission in ENDF Format |
| nesc1115 |
VISA-2, Reactor Vessel Failure Probability Under Thermal Shock |
| nesc9846 |
VISCOT, Viscous Mechanical Behaviour of Rock Mass Under Thermal Stress |
| iaea1324 |
VITEK, Non Stationary Navier-Stokes Solver for Compressible, Turbulent Flow |
| nea-0636 |
VIWI, Neutron Speeds and Weights for Scattering Kernel Calculation |
| nesc0922 |
VMCON, Minimization of Nonlinear Function with Constraints |
| ests0426 |
VODE, Variable Coefficient Ordinary Differential Equations (ODE) Solver |
| iaea0871 |
VPI-NECM, Nuclear Engineering Program Collection for College Training |
| nea-0655 |
VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation |
| iaea1417 |
W-SHIELDER, calculates shielding thickness of water for photon emitting radionuclide between 0.5 to 10 MeV |
| nea-1142 |
WADOSE, Radiation Source in Vitrification Waste Storage Apparatus |
| nea-0506 |
WAKE, Navier Stokes Equation with 2-D Turbulence, Stream Function, Vorticity |
| nesc9673 |
WAPPA, Waste Package Performance Assessment |
| uscd1157 |
WATEQ4F, Aqueous Speciation Calculation of Natural Waters |
| iaea1210 |
WEDRO, Data Processing Routines for WIMS-D/4 WIMSE File |
| nea-0610 |
WEERIE, Radioactive Release from Reactor to Cooling Circuit and Atmosphere |
| ests0160 |
WELBORE, Transient Wellbore Fluid Flow Model |
| iaea0821 |
WELWING, Material Buckling for HWR with Annular Fuel Elements |
| ests1197 |
WFSFIT, Wilson-Fowler Spline Fit Algorithm |
| nesc0278 |
WHAM-6, Pressure and Velocity Transients in Fluid Pipes, Wave Superposition Method |
| nea-1147 |
WHATIF-AQ, Geochem Speciation and Saturation of Aqueous Solution |
| iaea1243 |
WILIT, Utility Program for WIMS Library Handling |
| iaea0946 |
WILMA, WIMS Nuclear Data Library Maintenance |
| nea-0329 |
WIMS, MultiGroup Reactor Lattice Calculation for Thermal Reactor and Fast Reactor |
| ccc-0698 |
WIMS-ANL 4.0, Deterministic Code System for Lattice Calculation |
| iaea0887 |
WIMSCORE, 2-Group Constant from WIMS-D/4 for Programs TDB, TRITON, CITATION |
| nea-1507 |
WIMSD5, Deterministic Multigroup Reactor Lattice Calculations |
| iaea1254 |
WINTER, Interactive WIMS Input Preparation |
| iaea1408 |
WLUP3.0, 69 and 172 Group Cross Section Libraries for WIMS |
| ccc-0427 |
WRAITH, Internal and External Doses from Atmospheric Release of Isotopes |
| iaea0897 |
X4ECS, ENDF/B-4 and EXFOR Data Comparison |
| iaea0896 |
X4R, EXFOR Evaluated Data Retrieval |
| iaea0936 |
X4TOC4, Neutron Cross-Sections Data Conversion from EXFOR to Computation Format |
| nea-0564 |
XBWR, 1-D Xe Transients for BWR in Axial Geometry |
| nesc0988 |
XERROR, FORTRAN Library Error Message Processing Routines |
| nesc0572 |
XLACS, Fast Resonance and Thermal MultiGroup Cross-Sections from ENDF/B, Breit-Wigner, for Program XSDRN |
| iaea1395 |
XNWLUP, Graphical user interface to plot WIMS-D library multigroup cross sections |
| nesc0964 |
XOQDOQ, Meteorological Evaluation of Atmospheric Nuclear Power Plant Effluents |
| ccc-0525 |
XRAY-AAC, X-Ray Attenuation and Absorption |
| nesc0393 |
XSDRN, MultiGroup Cross-Sections from Resonance Data Library, Neutron Spectra and Group Constant Collapsing |
| nea-0072 |
ZADOC, 2 Group Time-Dependent Burnup in X-Y Geometry with Fuel Management |
| nea-0283 |
ZEUS-ALB.5, 3-D 1 Group Neutron Transport Kinetics in Slits, Channels, Tunnels by Monte-Carlo |
| nea-0401 |
ZOCO-6, Temperature Transients in BWR and PWR Containment During LOCA |
| nesc0765 |
ZONE, Finite Elements Method Quadrilateral and Triangular Mesh Generator for 2-D Axisymmetric Geometry |
| iaea1371 |
ZOTT99, Data Evaluation Using Partitioned Least-Squares |
| nea-0331 |
ZUBOK-2-3, Stability Region of Nonlinear 1st Order Differential Equation System by Lie-Series |
| nesc0041 |
ZUT, Resonance Integrals in Resolved Region at Various Temperature, Escape Probability Calculation |
| nea-1251 |
ZYLIND, Gamma Penetration for Cylindrical Source and Shield Geometry |
| nea-1398 |
ZZ 3DLWRCT, 3-D LWR Rod Ejection and Rod Withdrawal Benchmarks |
| nea-0789 |
ZZ ABBN, 26 Group Cross-Sections and Self Shielding Factors for Fast Reactors |
| nea-0790 |
ZZ ACTINIDES, 84-Group Neutron Cross-Section Library for Pu242 to Es253 Isotope Production Chain |
| iaea1275 |
ZZ ACTIV-87, Fast Neutron Activation Cross-Section |
| dlc-0069 |
ZZ ACTL82, Data Library of Evaluated Activation Cross-Sections |
| iaea1420 |
ZZ ADS-LIB/V1.0, test library for Accelerator Driven Systems v.1.0 |
| iaea1420 |
ZZ ADS-LIB/V2.0, test library for Accelerator Driven Systems v.2.0 |
| dlc-0049 |
ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section |
| dlc-0224 |
ZZ ALBEDO-DATA, Data for the Calculation of Albedos from Concrete, Iron, Lead and Water for Photons and Neutrons |
| nea-1745 |
ZZ ALEPH-LIB-JEFF3.1, MCNP Neutron Cross Section Library based on JEFF3.1 |
| nea-0886 |
ZZ AMPX-2/123, 123-Group Neutron Cross-Section Library from ENDF/B-4 by AMPX-2 |
| nea-0886 |
ZZ AMPX-2/219, 219-Group Neutron Cross-Section Library from ENDF/B-4 by AMPX-2 |
| dlc-0027 |
ZZ AMPX01/27C, Coupled Neutron-Gamma Group Constant Library by AMPX for Transport Calculation |
| iaea0912 |
ZZ AMZ, 70-Group 40 Isotope Multigroup Library for Fast Reactor Calculation |
| dlc-0129 |
ZZ ANS643, Geometric Progression Gamma-Ray Buildup Factor Coefficient Library |
| dlc-0154 |
ZZ ANSLV, Multigroup Cross Sections Library for ANS Reactor Design Studies |
| nea-0673 |
ZZ BABEL, Multigroup Neutron Cross-Section Data Library for Fast Reactor Shield Calculation |
| iaea0856 |
ZZ BARC-27GRP, 27-Group Infinitely Dilute and Bondarenko Cross-Section Library from ENDF/B |
| iaea1237 |
ZZ BARC-75, Coupled 50 Neutron-Group 25 Gamma-Group Cross-Section Library for 42 Nuclides |
| nea-1731 |
ZZ BFBT, OECD/NEA-US/NRC NUPEC BWR Full-size Fine-mesh Bundle Tests Benchmark |
| nea-1429 |
ZZ BIB-PU-RECY, Pu Recycling Bibliography |
| iaea1398 |
ZZ BOREHOLE-EB6.8-MG, multi group cross-section library for deterministic and Monte Carlo codes |
| dlc-0008 |
ZZ BP-3, 104-Group Neutron Cross-Section Library for Transport Calculation |
| dlc-0008 |
ZZ BP-6, 104 Group Neutron and Gamma-Ray Multigroup Cross-Section Library for Transport Calculation |
| iaea0949 |
ZZ BROND, Evaluated Neutron Data Library in ENDF-6 Format |
| nea-1401 |
ZZ BUC-1/BENCHMARK, NEACRP Benchmark Specifications for Burnup Criticality Calculation |
| nea-1866 |
ZZ BUGJEFF311.BOLIB, JEFF-3.1.1 Broad-Group Coupled X Sect Lib. for LWR Shielding & Pressure Vessel Dosimetry Applic. |
| dlc-0185 |
ZZ BUGLE-96, Multigroup Coupled Neutron Gamma Cross-Section for LWR Shielding Calculation |
| nea-1551 |
ZZ BWRSB-FORSMARKS, Stability Benchmark Data from BWR FORSMARKS 1 and 2 |
| nea-1454 |
ZZ BWRSB-RINGHALS1, Stability Benchmark Data from BWR RINGHALS-1 |
| nea-1640 |
ZZ BWRTT, BWR Turbine Trip Transient Benchmark Based on Peach-Bottom 2 |
| dlc-0059 |
ZZ CAD, 51 Neutron-Group, 25 Gamma-Group Albedo Data for 4 Materials from DOT Flux |
| dlc-0210 |
ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for CANDU Reactor Fuels |
| iaea1256 |
ZZ CENPL, Chinese Evaluated Nuclear Parameter Library |
| iaea1256 |
ZZ CENPL-DLS, Discrete Level Schemes and Gamma Branching Ratios Library |
| iaea1256 |
ZZ CENPL-FBP, Fission Barrier Paramater Library |
| iaea1256 |
ZZ CENPL-GDRP, Giant Dispole Resonance Parameter Library |
| iaea1256 |
ZZ CENPL-MCC, Nuclear Ground State Atomic Masses Library |
| iaea1256 |
ZZ CENPL-NLD, Nuclear Level Density Parameter Library |
| iaea1256 |
ZZ CENPL-OMP, Optical Model Parameter Library |
| iaea1297 |
ZZ CL50G, 50-Group Multigroup Library in AMPX Format for Fast Reactor Calculation |
| dlc-0042 |
ZZ CLEAR/42B, 126 Neutron-Group, 36 Gamma-Group Coupled Cross-Section in AMPX, CCCC Format, for LMFBR |
| nea-1775 |
ZZ CLES, cross section library of moderator materials for low-energy neutron sources |
| nea-1730 |
ZZ COV-15GROUP-2006, 15-group cross section covariance matrix library |
| dlc-0077 |
ZZ COVERV, Multigroup Cross-Section Covariance Matrices in COVERX Format |
| dlc-0091 |
ZZ COVFILS, 30-Group Covariance Library from ENDF/B-5 for Sensitivity Studies |
| dlc-0137 |
ZZ COVFILS-2, 74-Group Neutron Cross-Section, Scattering Matrices, Covariances for Fusion Reactors |
| dlc-0138 |
ZZ COVFILS-2-I, 74-Group Neutron Cross-Section, Scattering Matrices, Covariances for Fusion Reactors |
| nea-1787 |
ZZ CRYO-S(A,B)-ACE1, Scattering law and continuous energy cross section library of materials at cryogenic temperatures |
| dlc-0130 |
ZZ DABL69, 46-Group Neutron, 23-Group Gamma Cross-Section in ANISN Format from ENDF/B-V |
| nea-0791 |
ZZ DAMSIG84, 640-Group Damage Cross-Section Library for SAND-2 Calculation |
| dlc-0030 |
ZZ DECAYREM/C, Decay Spectra Library for EXREM Calculation |
| nea-1644 |
ZZ DECDC, Nuclear Decay Data Files for Dose Calculation |
| nea-1538 |
ZZ DECNET-GENDF, Fusion Damage Library of 175 Neutron and 42 Photon VITAMIN-J Groups |
| dlc-0010 |
ZZ DLC-10B AVKER, Neutron Kerma Response Function Data Library |
| dlc-0011 |
ZZ DLC-11 RITTS, 121-Group Coupled Cross-Section for ANISN, DOT, MORSE |
| dlc-0012 |
ZZ DLC-12D POPLIB, Secondary Gamma Yields and Cross-Section Library for POPOP-4 Calculation |
| dlc-0013 |
ZZ DLC-13B, Resonance Cross-Section Group Constant Library for Tungsten and Depleted Pu |
| dlc-0014 |
ZZ DLC-14 AIR, Group Constant Library of Secondary Gamma Transport in Air for ANISN Calculation |
| dlc-0015 |
ZZ DLC-15 STORM-ISRAEL, Gamma Interaction Cross-Section Library in ENDF/B Format for Transport |
| dlc-0016 |
ZZ DLC-16 COBB, 123 Neutron-Group Cross-Section Library from ENDF/B for XSDRN Calculation |
| dlc-0017 |
ZZ DLC-17 NOX, 119-Group Coupled Cross-Section of Nitrogen, Oxygen, Air for MORSE |
| dlc-0018 |
ZZ DLC-18 NAB, 100 Neutron-Group Cross-Section Library of Na and Al for ANISN, DOT, MORSE Neutron Transport |
| dlc-0019 |
ZZ DLC-19 DECAYGAM, Isotope Gamma Energy Library for Spectrometry Evaluation |
| dlc-0021 |
ZZ DLC-21, X-Ray Attenuation Cross-Section Library from 0.1 KeV to 1 MeV |
| dlc-0023 |
ZZ DLC-23F CASK, 40-Group Neutron and Gamma Coupled Cross-Section for PWR Shipping Casks |
| dlc-0028 |
ZZ DLC-28, 73-Group Neutron and Gamma Coupled Cross-Section for CTR Transport Calculation |
| dlc-0002 |
ZZ DLC-2D/100G, 100 Neutron-Group Cross-Section Library by SUPERTOG Calculation for ANISN, DOT |
| dlc-0031 |
ZZ DLC-31, 37 Neutron-Group, 21 Gamma-Group Coupled Group Constants Library from ENDF/B |
| dlc-0006 |
ZZ DLC-6 GAMLIB, 99-Group Cross-Section Library by SUPERTOG Calculation from ENDF/B |
| dlc-0090 |
ZZ DOSCOV, 24-Group Covariance Data Library from ENDF/B-V for Dosimetry Calculation |
| nea-0827 |
ZZ DOSCROS, Neutron Cross-Section Library for Spectra Unfolding and Integral Parameter Evaluation |
| dlc-0079 |
ZZ DOSDAT-2, Gamma and Electron Dose Conversion Factor Data Library for Body Organs |
| dlc-0144 |
ZZ DOSEDAT-DOE, Dose-Rate Conversion Factors for External Photon, Electron Exposure |
| dlc-0080 |
ZZ DRALIST, Radioactive Decay Data for Dosimetry and Hazard Assessment |
| iaea1401 |
ZZ DROSG-2000, Legendre Coefficient Library for 59 monoenergetic neutron source reactions |
| nea-1609 |
ZZ EAF 99, Cross Section Library for Neutron Induced Activation Materials |
| nea-1606 |
ZZ ECN-BUBEBO, ECN-Petten Burnup Benchmark Book, Inventories, Afterheat |
| dlc-0106 |
ZZ ECPL86, Data Library of Evaluated Charged Particle Cross-Section, Nuclides Up to Oxygen |
| nea-1050 |
ZZ EFF1LIB, Fusion Fast Neutron Data Library for MCNP |
| dlc-0208 |
ZZ ELAST2, Database of Cross Sections for the Elastic Scattering of Electrons and Positrons by Atoms |
| dlc-0100 |
ZZ ELECSPEC, Electron Spectra Data Library from Fission Product Decay |
| uscd0803 |
ZZ ENDF/B-IV, Evaluated Nuclear Data File Version 4 |
| uscd1233 |
ZZ ENDF/B-V, Evaluated Nuclear Data File Version 5 |
| dlc-0103 |
ZZ ENDL86, Evaluated Charged Particle, Neutron, Photon Cross-Section Library |
| dlc-0179 |
ZZ ENDLIB, Coupled Electron and Photon Transport Library in ENDL Format |
| dlc-0037 |
ZZ EPR/37F, 100 Neutron-Group, 21 Gamma-Group Coupled Cross-Section for Experimental Power Reactor (EPR) Fusion System |
| 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 |
| nea-1424 |
ZZ FSXLIBJ33, MCNP nuclear data library based on JENDL-3.3 |
| nesc0844 |
ZZ FUELS-DATA, Data Library for LWR Fuel Behaviour for FRAP Program |
| nea-0878 |
ZZ GAMDAT-78, Gamma Decay Data of Radioisotopes |
| dlc-0071 |
ZZ GAMMON, Activation Data Library for Fusion Reaction |
| nea-1543 |
ZZ GEFF-2-GENDF, P5 175-N and P8 42-Gamma Group Library for Fusion Blanket Applications |
| nea-1544 |
ZZ GEFF-2-MATXS, Coupled Neutron-Gamma Fusion Neutronics Library in MATXS Format |
| nea-1102 |
ZZ GEFF1, 175-Group Neutron Cross-Section in VITAMIN-J1 Format for Shielding Benchmarks |
| nea-1255 |
ZZ GREAC-ECN-3 REAC-ECN-4, Neutron Reaction Cross-Sections Library for Fusion Reactors |
| nea-1344 |
ZZ GROUPSTRUCTURES, VITAMIN-J, XMAS, ECCO-33, ECCO2000 Standard Group Structures |
| nea-1210 |
ZZ HATCHES-19, Database for radiochemical modelling |
| dlc-0220 |
ZZ HILO2K, Coupled 83 Neutron, 22 Photon Group Cross Sections for Neutron Energies Up to 2 GeV for ANISN, DORT and TORT |
| dlc-0119 |
ZZ HILO86, 66 Neutron, 22 Gamma Group Cross-Section Library for ANISN, DORT, MORSE |
| dlc-0187 |
ZZ HILO86R, 66 Neutron, 22 Gamma Group Cross-Section for 400 MeV Neutron, 20 MeV Gamma |
| dlc-0007 |
ZZ HPICE/F, Gamma Interaction Cross-Section Library in ENDF/B Format for Transport Calculation |
| dlc-0099 |
ZZ HUGO, Photon Interaction Data Library in ENDF-5 Format |
| dlc-0146 |
ZZ HUGO-VI, Photon Interaction Data in ENDF-6 Format |
| iaea1419 |
ZZ IBANDL, Ion Beam Analysis Nuclear Data Library in R33 format |
| nea-1656 |
ZZ IEAF-2001, Intermediate Energy Activation File |
| iaea1418 |
ZZ INDL/TSL, Thermal Neutron Scattering Data for H2O, D2O and ZrHx in ENDF-6 Format and as MCNP(X) Data Sets |
| iaea1215 |
ZZ IRAN-LIB, Multigroup Neutron Gamma Cross-Section Library for 33 Elements in ANISN Format |
| iaea0867 |
ZZ IRDF-2002, 640-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-6 Format |
| iaea0867 |
ZZ IRDF-2002, 640-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-6 Format |
| iaea0867 |
ZZ IRDF-2002-ACE, Cross-Section Library and Spectra for Dosimetry Calculation in ACE Format for Monte Carlo methods |
| iaea0867 |
ZZ IRDF-2002-ACE, Cross-Section Library and Spectra for Dosimetry Calculation in ACE Format for Monte Carlo methods |
| iaea0867 |
ZZ IRDF-82, 620-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-5 Format |
| iaea0867 |
ZZ IRDF-82, 620-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-5 Format |
| iaea0867 |
ZZ IRDF-90, 640-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-6 Format |
| iaea0867 |
ZZ IRDF-90, 640-Group Cross-Section Library and Spectra for Dosimetry Calculation in ENDF-6 Format |
| nea-1853 |
ZZ JENDL-1, Japanese Evaluated Nuclear Data Library |
| nea-1624 |
ZZ JENDL/D-99, JENDL Dosimetry Cross-Sections Data Library and Graphical Representations |
| nea-0796 |
ZZ JFS-1, Cross-Sections Library 25-Groups ABBN and 70-Group JFS for Fast Reactor Calculation |
| nea-0796 |
ZZ JFS-2, 25 Group (ABBN) and 70 Group JFS Cross Sections Library for Fast Reactors |
| nea-0796 |
ZZ JFS-3/J2, 70 Group 30 Isotopes Cross Section Library for Fast Reactors |
| nea-1815 |
ZZ KAFAX-E70, 150 and 12 Groups Cross Section Library in MATXS Format based on ENDF/B-VII.0 for Fast Reactors |
| nea-1650 |
ZZ KAFAX-F22, 80 and 24 Groups Cross-Section Library in MATXS Format Based on JEF-2.2 for Fast Reactors |
| nea-1816 |
ZZ KAFAX-F31, 150 and 12 Groups Cross Section Library in MATXS Format based on JEFF-3.1 for Fast Reactors |
| nea-1817 |
ZZ KAFAX-J33, 150 and 12 Groups Cross Section Library in MATXS Format based on JENDL-3.3 for Fast Reactors |
| nea-1848 |
ZZ KALININ3, KALININ-3 Coolant Transient Benchmark |
| dlc-0160 |
ZZ KAOS/LIB-V, Kerma Factors, Nuclear Response Function Library for Fission, Fusion |
| nea-1649 |
ZZ KASHIL-E6, 175 N, 42 Gamma Groups Cross Sections in MATXS Format Based on ENDF/B-VI.5 for Shielding Applications |
| nea-1818 |
ZZ KASHIL-E70, 199 N, 42 Photon Groups Cross Sections in MATXS Format Based on ENDF/B-VII.0 for Shielding Applications |
| dlc-0142 |
ZZ KERMAL, Neutron and Gamma Kerma Library from ENDL and EGDL |
| iaea0870 |
ZZ L26P3S34, 26-Group Constants Library of 34 Materials for Neutron Shielding Calculations |
| dlc-0168 |
ZZ LA100, ENDF Format Data Library for Neutron and Protons Up to 100 MeV |
| dlc-0054 |
ZZ LAFPX-V, Multigroup Fission Product Data Library from ENDF/B-V by Program NJOY |
| dlc-0128 |
ZZ LAHIMACK, Multigroup Neutron and Gamma Cross-Section and Response Function up to 800 MeV |
| nesc0532 |
ZZ LASL-XSECS, Fast and Thermal Multigroup Cross-Section Library in LANL Transport Format |
| dlc-0040 |
ZZ LIB-IV, 50-Group Cross-Section Library in CCCC-III Format from ENDF/B-IV for Fast Reactors |
| dlc-0089 |
ZZ LUMP, Lumped Fission Product Cross-Section Library for Fast Reactor Analysis from ENDF/B-V |
| dlc-0029 |
ZZ MACKLIB, Nuclear Response Function Library for CTR and Hybrid Fission Fusion System Materials |
| dlc-0060 |
ZZ MACKLIB-4, 171-Neutron, 36-Gamma Group Response Function Library from ENDF/B-IV |
| nea-1740 |
ZZ MATJEF22.BOLIB, JEF-2.2 Multigr Coupled (199n + 42gamma) X-Section Lib. in MATXS Fmt for Nuclear Fission Applications |
| nea-1847 |
ZZ MATJEFF31.BOLIB, JEFF-3.1 Multigr Coupled(199n + 42gamma) X-Section Lib.in MATXS Fmt for Nuclear Fission Applications |
| nea-1205 |
ZZ MATX175/42-JEFF87, 172 Neutron-Group, 42 Gamma-Group MATXS Library in VITAMIN-J Structure |
| dlc-0176 |
ZZ MATXS10, 30-Group Neutron, 12-Group Gamma Cross-Sections in MATXS Format from ENDF/B-VI |
| dlc-0177 |
ZZ MATXS11, 80-Group Neutron, 24-Group Gamma Cross-Section in MATXS Format from ENDF/B-VI |
| nea-1206 |
ZZ MATXS70-JEFF87, 69+1 Group MATXS Library in WIMS BOXER Structure |
| nea-1707 |
ZZ MATXSLIBJ33,JENDL-3.3 based,175 N-42 photon groups (VITAMIN-J) MATXS lib. for discrete ordinates multi-group |
| nea-1668 |
ZZ MCB-EAF99, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K. |
| nea-1669 |
ZZ MCB-ENDF/B6.8, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K. |
| nea-1667 |
ZZ MCB-JEF2.2, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K. |
| nea-1670 |
ZZ MCB-JENDL-3.2, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K. |
| nea-1655 |
ZZ MCB63NEA.BOLIB, MCNP Cross Section Library Based on ENDF/B-VI Release 3 |
| nea-1616 |
ZZ MCJEF22NEA.BOLIB, MCNP Cross Section Library Based on JEF-2.2 |
| nea-1768 |
ZZ MCJEFF3.1NEA, MCNP Neutron Cross Section Library based on JEFF3.1 |
| nea-1651 |
ZZ MCLIB-E6, Continuous Energy Cross Section Library from ENDF/B-VI.5 for MCNP-4A, -4B, 300K, 600K, 900K |
| dlc-0200 |
ZZ MCNPDATA, ZZ-MCB-DLC200, Standard Neutron, Photon and Electron Data Libraries for MCNP-4C and MCB1C |
| iaea1376 |
ZZ MENDL-2P, Proton Medium Energy Nuclear Data Library |
| nea-1613 |
ZZ MICROX 2 FSS LIB, Data Library for Fast Spectrum Systems Analysis |
| iaea1412 |
ZZ MINSKACT, Evaluated neutron reaction data for Th-232, Pa, U, Np, Pu, Am and Cm isotopes |
| dlc-0033 |
ZZ MONTAGE-400, Neutron Activation 100-Group Cross-Section Library of Fusion Reactor Materials |
| iaea1217 |
ZZ N-SPECT/DET-RESP, Neutron Spectra and Detector Responses for Radiation Protection |
| iaea1279 |
ZZ NMF-90, Database for Neutron Spectra Unfolding |
| dlc-0172 |
ZZ NUCDECAY, Nuclear Decay Data for Radiation Dosimetry Calculation for ICRP and MIRD |
| dlc-0202 |
ZZ NUCDECAYCALC, Nuclear Decay Data for Radiation Dosimetry Calculation for ICRP |
| nea-1618 |
ZZ ORESUND, Nordic Mesoscale Dispersion Experiments over Land-Water-Land |
| nea-1642 |
ZZ ORIGEN2.2-UPJ, A complete package of ORIGEN2 libraries based on JENDL-3.2 and JENDL-3.3. |
| dlc-0038 |
ZZ ORYX-E/38B, Group Constant Library from ENDF/B Fission Product Data for ORIGEN Calculation |
| iaea1423 |
ZZ PADF-2007, Proton Activation Data File in ENDF-6 format |
| nea-1746 |
ZZ PBMR-400, OECD/NEA PBMR Coupled Neutronics/Thermal Hydraulics Transient Benchmark - The PBMR-400 Core Design |
| dlc-0236 |
ZZ PHOBIA, Photon Buildup Factors to Account for Angular Incidence on Shield Walls |
| dlc-0136 |
ZZ PHOTX, Photon Interaction Cross-Section Library for 100 Elements |
| nea-1868 |
ZZ PIXE2010, Proton/alpha ionization (K,L,M shell) tabulated cross section library |
| iaea1235 |
ZZ PNESD, Diffusion Elastic Scattering Cross-Section of 3 MeV to 1000 MeV Proton on Natural Isotopes |
| iaea1409 |
ZZ POINT-2004, Linearly Interpolable ENDF/B-VI.8 Data for 13 Temperatures |
| iaea1421 |
ZZ POINT-2007, linearly interpolable ENDF/B-VII.0 data for 14 temperatures |
| iaea1430 |
ZZ POINT-2009, a Temperature Dependent ENDF/B-VII.0 Cross Section Library |
| dlc-0212 |
ZZ POINT2000, Linearly Interpolable ENDF/B-VI.7 Data for 8 Temperatures |
| dlc-0247 |
ZZ POINT2011, linearly interpolable ENDF/B-VII.1 Beta2 cross section library for 13 temperatures |
| dlc-0192 |
ZZ POINT97, Temperature-Dependent ENDF/B-6 Cross-Sections at 8 Temperature Between 0K and 2100K |
| dlc-0196 |
ZZ PR-EDB, Power Reactor Embrittlement Database |
| nea-1849 |
ZZ PSBT, NUPEC PWR Sub-channel Bundle Tests Benchmark |
| dlc-0126 |
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