catalog | categories | new | search |

Catalog of Programs in Category C

C. Static Design Studies


nesc0325 2-DB, 2-D MultiGroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search
nea-0108 ALCI, Homogeneous 2-D MultiGroup Neutron Diffusion in X-Z, R-Z, R-Theta Geometry with Criticality Search
ccc-0558 ALKASYS, Rankine-Cycle Space Nuclear Power System
psr-0315 AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5
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-0546 APPLE, Plot of 1-D MultiGroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN
nea-0320 ARGO, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry from JAERI Fast-Set, ABBN, RCBN
nea-0661 ASNBILD, Generator of JCL and Data for Program ANISN on CDC Computer
ccc-0519 AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors
nea-0179 AXIFLUX, Cosine Function Fit of Experimental Axial Flux in Cylindrical Reactor
ccc-0459 BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup
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
iaea1190 BRETISLAV,OLDRICH, Neutron Diffusion in Hexagonal Geometry for WWER Critical Fuel Assemblies
nesc0270 CAESAR-4, 1-D MultiGroup Diffusion in Slab, Cylindrical, Slab Geometry, Criticality Search
nea-0649 CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback
iaea0920 CEBIS, 1-D 2 Group Diffusion Code for Reactor Calculation
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
nea-0357 CLUP-77C, Collision Probability and Neutron Flux Distribution in Multi-Region Fuel Cluster
ccc-0726 CNCSN 2009, One, Two- and Three-Dimensional Coupled Neutral and Charged Particle Sn Parallel Multi-Threaded Code System
ccc-0777 COG 11, Multiparticle Monte Carlo Code System for Shielding and Criticality Use
iaea1226 CORD, PWR Core Design and Fuel Management
nea-0057 CRAM-360, 1-D, 2-D Multi Group Diffusion with Keff Calculation or Criticality Search
nea-1416 D3D/D3E, 2-D, 3-D MultiGroup Neutron Diffusion in Rectangular, Cylindrical, Triangular-Z Geometry
nea-0672 DIAMANT-2, MultiGroup Neutron Transport with Anisotropic Scattering in Triangular Geometry
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
nea-0808 DIFFUSION-ACE, 3-D Neutron Diffusion by Leakage Iteration Method
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
ccc-0650 DOORS, 1-,2-,3-dimensional discrete-ordinates system for deep-penetration neutron and photon transport
ccc-0543 DORT, 1-D 2-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration
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
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
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-1683 ERANOS 2.0, Modular code and data system for fast reactor neutronics analyses
nea-0534 EREBUS, Burnup by 2-D MultiGroup Neutron Diffusion with Criticality Search
nea-0449 ESTOQ, 1st Collision Source Calculation for Program DOT in R-Z Geometry
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
iaea0835 FASVER, 2 Group 2-D Diffusion in X-Y Geometry and Adjoint Solution for PWR with Reflector
psr-0563 FEAST-METAL-V.1.0, Fuel Engineering and Structural analysis Tool
nea-0443 FEM-2D, 2-D MultiGroup Diffusion in X-Y Geometry
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-0478 FEMB, 2-D Homogeneous Neutron Diffusion in X-Y Geometry with Keff Calculation, Dyadic Fission Matrix
ests0486 FESH, 2-D Multigroup Neutron Transport with Isotropic Scattering
iaea1221 FIGA, Source Distribution Conversion from X-Y to R-Theta Geometry
nea-0896 FINELM, MultiGroup Diffusion in 3-D by Finite Elements Method
nesc0167 FLARE, 3-D BWR Reactivity and Power Distribution Appraisal Calculation
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
ccc-0603 FPZD, Reactor Burnup by MultiGroup Neutron Diffusion
nea-1021 FURNACE, Neutronic Calculation in 3-D Toroidal Geometry
nesc0606 GAPER-1D, 1-D MultiGroup 1st Order Perturbation Transport Theory for Reactivity Coefficient
nesc0380 GATT, 3-D Few-Group Neutron Diffusion for Power Distribution in Hexagonal Reactor Core for HTGR
nea-0605 GENP-2, Program System for Integral Reactor Perturbation
psr-0304 GIRAFFE, Isotope Release Analysis in LMFBR Fuel Elements Failure
iaea1271 GNOMER, Core Power Distribution by 1-D, 2-D, 3-D MultiGroup Neutron Diffusion
iaea0908 GRENADE, Green's Function Nodal Algorithm for Diffusion Equation
ccc-0276 GRUNCLE, 1st Collision Source Calculation for Program DOT
nesc0277 HAMMER, 1-D MultiGroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation
nesc0136 HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor
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
iaea0914 HEXNOD23, 2-D, 3-D Coarse Mesh Solution of Steady State Diffusion Equation in Hexagonal Geometry
nea-1871 JN-METD, N Transport with Isotropic Scattering, Bare Slabs and Homogeneous Slabs (JN-METD1), Multilayer Slabs (JN-METD2)
nea-0624 JOSHUA, Neutronics, Hydraulics, Burnup, Refuelling of LWR
nea-0343 KASY, 3-D Homogeneous Neutron Diffusion in X-Y-Z, R-Theta, Hexagonal-Z Geometry by Synthesis Method
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
ccc-0548 KENO5A-PC, Monte-Carlo Criticality with Supergrouping
nea-0616 KIM, Steady-State Transport for Fixed Source in 2-D Thermal Reactor by Monte-Carlo
iaea1232 LABAN-PEL, 2-D MultiGroup Neutron Diffusion in X-Y Geometry by Response Matrix Eigenvalue
nea-0167 LIE-PN, Pn Neutron Transport in Radial Geometry Cell with Source Problems Calculation
nea-0836 MADONNA, Neutron Flux with Void Region by Removal Diffusion Method
nea-0528 MARC, MultiGroup Diffusion in R-Z, X-Y, R-Theta, Slab, Cylindrical, Spherical, Hexagonal, Triangular Geometry
nea-0926 MARIA-SYSTEM, PWR Fuel Assembly Calculation for Program CARMEN-System and Simulation
uscd1241 MCART, solve the time dependent neutron transport equation
nea-1643 MCB1C, Monte-Carlo Continuous Energy Burnup Code
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-1005 MOBIDIC, Fast Reactor Hexagonal Infinite Lattice 2 Component Fuel Pin Diffusion Coefficient
iaea1238 MOCA, Criticality of VVER Reactor Hexagonal Fuel Assemblies
nea-1279 MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors
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
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
nea-1633 MOSRA-LIGHT, High Speed 3-D X-Y-Z Nodal Diffusion Code for Vector Computers
nea-0933 MULTI-KENO, Criticality Safety Analysis by Monte-Carlo
nea-0035 MUPO, Critical 43 Group Spectra Calculation for Homogeneous Reactor
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-1673 MVP/GMVP II, MC Codes for Neutron & Photon Transport Calc. based on Continuous Energy and Multigroup Methods
iaea1173 NEHEX-3D, 3-D Neutron Diffusion for Fast Reactors and WWER in Hexagonal Geometry
ccc-0641 NESTLE, Few-Group Neutron Diffusion for Steady-State and Transient Problems by Nodal Expansion Method (NEM)
psr-0355 NJOY-94, General ENDF/B Processing System for Reactor Design Problems
psr-0171 NJOY91, General ENDF/B Processing System for Reactor Design Problems
iaea1171 NOTRAN/3D, 3-D Neutron Transport in X-Y-Z Geometry by Discrete Nodal Transport Method
nea-1591 OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo
ccc-0266 ONETRAN, 1-D Transport in Planar, Cylindrical, Spherical Geom. for Homogeneous, Inhomogeneous Probl., Anisotropic Source
nea-1324 OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN
nea-0702 PALLAS-2DY, 2-D Neutron Transport, Gamma Transport with Anisotropic Scattering for Fixed Source
ccc-0760 PARTISN 5.97, 1-D, 2-D, 3-D Time-Dependent, Multigroup Deterministic Parallel Neutral Particle Transport Code
nea-1238 PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation
nea-0464 PN, 1-D, 2-D, 3-D MultiGroup Neutron Transport
nea-0181 RADIFLUX, Bessel Function Fit of Radial Flux Data in Cylindrical Reactor
nesc0631 RAFFLE-2/MOD2, 3-D Steady-State Monte-Carlo Neutron Transport, Cell Averaged Scattering Cross-Sections Calculation
ccc-0708 REBUS-PC 1.4, Code System for Analysis of Research Reactor Fuel Cycles
nea-0262 REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR
iaea0929 RICANT, Neutron Transport in X-Y Geometry for Isotropic Scattering
nea-0598 RSYST, Modular System for Reactor Core and Shielding Problems
nea-1779 SAGEP-FR, Sensitivity Analysis of Fast Reactor Parameters
ccc-0785 SCALE 6.1.2, Modular system for criticality, shielding, source term, fuel depletion/decay, inventories, reactor physics
nea-1840 SERPENT 1.1.7, 3-D continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications
nea-0370 SHADOK-3-6, Transport Equation with Anisotropic Diffusion in P1 Approximation for Spherical and Cylindrical Geometry
nea-0319 SIMPLED-4, 1-D Neutron Diffusion in Spherical, Cylindrical, Planar Geometry from JAERI Fast-Set and ABBN
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-1081 SLAROM, Neutron Flux Distribution and Spectra in Lattice Cell
nea-0430 SNAP, MultiGroup 3-D Neutron Diffusion in X-Z, R-Theta-Z, Hexagonal-Z, Triangular-Z Geometry
nea-1826 SOLTRAN, solving multi-dimensional simplified P2 transport and diffusion problems of hexagonal geometry in fast reactors
nea-0414 SQUID-360, 2-D Neutron Diffusion in X-Y and R-Z Geometry with Criticality Search and Constant Neutron Source
nea-0703 STEADY-ACE, 3-D Neutronics and Multichannel Thermohydraulics Analysis of BWR
ccc-0248 SWAN-PPL, Fusion Reactor 1-D Particle Transport Optimization
ccc-0204 SWANLAKE, Cross-Sections Sensitivity Analysis for 1-D Discrete Ordinate Calculation
nesc0713 SYN-3D, 2-D and 3-D Neutron Diffusion Static Eigenvalues, Single Channel Spatial Flux Synthesis
iaea1383 SYRCO-1, 1-D, 2 Groups Multi-Zone Interactive Diffusion Code
ccc-0638 TART2012, 3D Coupled Neutron-Photon Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code
nesc0558 TASK, 1-D MultiGroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron
nea-0997 THT, 3-D Coarse Mesh LWR Bundle Fluxes and Power with Discontinuity Factors
ccc-0759 TITAN 1.29, A Three-Dimensional Deterministic Radiation Transport Code System
ccc-0543 TORT, 2-D 3-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration
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-0953 TRANSHEX, 2-D Thermal Neutron Flux Distribution from EpiThermal Flux in Hexagonal Geometry
nea-0117 TRAWS-4, Axial Flux Distribution for Control Rod Variations
ccc-0293 TRIDENT, 2-D Neutron Transport for Homogeneous and Inhomogeneous Problems in X-Z, R-Z Geometry, Anisotropic Scattering
nea-0384 TRIGON, 2-D Homogeneous and Inhomogeneous Fixed Source Neutron Diffusion for Triangular or Hexagonal Mesh
nea-1716 TRIPOLI-4 version 8.1, 3D general purpose continuous energy Monte Carlo Transport code
nea-1878 TRIPOLI-4 version 9S, Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo Transport Calculation
nea-1086 TRISTAN, 3-D fixed source radiation transport
nea-1087 TRITAC, 3-D Transport by Discrete Ordinate Method in X-Y-Z Geometry
nea-0415 TRITON, 3-D Multi-Region Neutron Diffusion Burnup with Criticality Search
nea-0471 TVEDIM, 2-D Homogeneous and Inhomogeneous Neutron Diffusion for X-Y, R-Z, R-Theta Geometry
ccc-0547 TWODANT-SYS, DANTSYS3.0, 1-D, 2-D, 3-D MultiGroup Discrete Ordinate Method Transport
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
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
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
nea-1856 VESTA 2.1.5, Monte Carlo depletion interface code and AURORA 1.0.0, Depletion analysis tool
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
iaea0871 VPI-NECM, Nuclear Engineering Program Collection for College Training
nea-0655 VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation
nea-1882 XSUN-2013, Windows interface environment for transport and sensitivity-uncertainty software TRANSX-2, PARTISN and SUSD3D
iaea1237 ZZ BARC-75, Coupled 50 Neutron-Group 25 Gamma-Group Cross-Section Library for 42 Nuclides
iaea0949 ZZ BROND, Evaluated Neutron Data Library in ENDF-6 Format
dlc-0042 ZZ CLEAR/42B, 126 Neutron-Group, 36 Gamma-Group Coupled Cross-Section in AMPX, CCCC Format, for LMFBR
dlc-0011 ZZ DLC-11 RITTS, 121-Group Coupled Cross-Section for ANISN, DOT, MORSE
dlc-0016 ZZ DLC-16 COBB, 123 Neutron-Group Cross-Section Library from ENDF/B for XSDRN Calculation
dlc-0018 ZZ DLC-18 NAB, 100 Neutron-Group Cross-Section Library of Na and Al for ANISN, DOT, MORSE Neutron Transport
dlc-0002 ZZ DLC-2D/100G, 100 Neutron-Group Cross-Section Library by SUPERTOG Calculation for ANISN, DOT
dlc-0006 ZZ DLC-6 GAMLIB, 99-Group Cross-Section Library by SUPERTOG Calculation from ENDF/B
dlc-0187 ZZ HILO86R, 66 Neutron, 22 Gamma Group Cross-Section for 400 MeV Neutron, 20 MeV Gamma
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
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
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
dlc-0076 ZZ SAILOR, 47 Neutron-20 Gamma-Group Coupled Cross-Section Library from VITAMIN-C by AMPX
dlc-0024 ZZ SINEX, 100 Neutron-Group Neutron Reaction Cross-Section Library from ENDF/B by SUPERTOG for ANISN
iaea0865 ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD
nea-1264 ZZ VITAMIN J/COVA, Covariance Matrix Data Library for Uncertainty Analysis
dlc-0041 ZZ VITAMIN-C/B, 171 Neutron-Group, 36 Gamma-Group Coupled Cross-Section for Fusion, LMFBR Calculations
nea-1207 ZZ WIMS-LIB/JEF87, 69+1 Group WIMS-D Library from JEF-1