MELCOR is a fully integrated, engineering-level computer code developed by
Sandia National Laboratories
Sandia National Laboratories (SNL), also known as Sandia, is one of three research and development laboratories of the United States Department of Energy's National Nuclear Security Administration (NNSA). Headquartered in Kirtland Air Force Ba ...
for the
U.S. Nuclear Regulatory Commission
The Nuclear Regulatory Commission (NRC) is an independent agency of the United States government tasked with protecting public health and safety related to nuclear energy. Established by the Energy Reorganization Act of 1974, the NRC began operat ...
to model the progression of
severe accidents in
nuclear power plants
A nuclear power plant (NPP) is a thermal power station in which the heat source is a nuclear reactor. As is typical of thermal power stations, heat is used to generate steam that drives a steam turbine connected to a generator that produces elec ...
. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. MELCOR applications include estimation of severe accident source terms, and their sensitivities and uncertainties in a variety of applications.
See also
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Nuclear engineering
Nuclear engineering is the branch of engineering concerned with the application of breaking down atomic nuclei ( fission) or of combining atomic nuclei (fusion), or with the application of other sub-atomic processes based on the principles of n ...
*
Monte Carlo method
Monte Carlo methods, or Monte Carlo experiments, are a broad class of computational algorithms that rely on repeated random sampling to obtain numerical results. The underlying concept is to use randomness to solve problems that might be determi ...
*
Nuclear reactor
A nuclear reactor is a device used to initiate and control a fission nuclear chain reaction or nuclear fusion reactions. Nuclear reactors are used at nuclear power plants for electricity generation and in nuclear marine propulsion. Heat from nu ...
*
MCNP
Monte Carlo N-Particle Transport (MCNP) is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation transport code designed to track many particle types over broad ranges of energies and is developed by L ...
External links
SNL MELCOR website*
Wikiversity: Nuclear Engineering
Nuclear safety and security
Physics software
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