CCC-0547
DANTSYS3.0
1-D, 2-D, 3-D, Multigroup Discrete Ordinates Methods for Radiation Transport
CCC-0704
SLIDERULE 1.0
Slide Rule for direct radiation exposure approximation in criticality accidents
CCC-0553
RASCAL 3.0.1
Radiological doses from accidental release to atmosphere
CCC-0638
TART2000
3D Coupled N-Photon Combinatorial Geometry Time Dependent Monte Carlo Transport Code
CCC-0684
NRCDose 2.3.2
Evaluation of Routine Radioactive Effluents from Nuclear Power Plants
DLC-0045
ZZ SENPRO/45C
multigroup sensitivity library for fast reactors thermal reactors
PSR-0226
PRECO-2000
Exciton Model Preequilibrium Code System with Direct Reactions
PSR-0137
MARLOWE 15a
Computer Simulation of Atomic Collisions in Crystalline Solids.
CCC-0365
Calculating the Estimation of Dose to the World Population from Releases of Iodine-129 to the Environment
PSR-0516
PARET-ANL
Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores
NEA-1645
IFPE/EFE-RO
Experimental Fuel Elements RO89 and RO51 in TRIGA 14 MW reactor (INR-Pitesti)
NEA-1622
IFPE/OSIRIS R1
4 PWR Rods Irradiated in the CEA Osiris Reactor
NEA-1536
IFPE/TRIBULATION R1
Fuel rod behaviour at high burnup Belgonucleaire & BBR
NEA-1625
IFPE/GAIN
gadolinia doped UO2 fuel behaviour experiment
NEA-1561
CHEMENGL/CHIMISTE
Chemical and Physical Properties of Elements
NEA-1492
NUCLEUS-CHART
Interactive chart of nuclides
IAEA1287
MC code for simulating interaction of high energy hadrons with complex macroscopic targets
NEA-1554
ZZ PWR-MSLB REV.2
PWR Main Steam-Line Break Benchmarks Coupled Neutronics Thermal Hydraulics
NEA-1657
Argonne National Laboratory Code Center Benchmark Problem Book
NEA-1581
ART MOD2
Fission Product Migration in Primary System and Containment
NEA-1313
Thermal Hydraulic Analysis of a BWR plant
NEA-1577
SKETCH-N 1.0
Solve Neutron Diffusion Equations of Steady-state and Kinetics Problems
NEA-1578
COMRAD96
Nuclear Fuel Burnup and Depletion Calculation
NEA-1593
TRAC-PF1/EN MOD 3
Best Estimate Coupled 3D Neutronics-thermalhydraulics
IAEA1240
HEXAB-3D
3 - D Few-group Diffusion for Hexagonal Core Geometry
NEA-1517
SINBAD-ASPIS-NG
ASPIS Neutron/Gamma-Ray Transport Through Water/Steel Arrays
NEA-1517
SINBAD-IRI-TUB-DUCT
IRI-TUB Streaming Through Ducts
NEA-1553
SINBAD-FNS-C-CYLIND
Integral Experiment on a 60 cm-thick Graphite Cylindrical Assembly (FNS/JAERI clean benchmark)
NEA-1553
SINBAD-OKTAVIAN/SI
Leakage Neutron and Gamma Spectra from 40 and 60 cm diameter Silicon Sphere Pile With 14 MeV Neutrons