MCNP
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What is MCNP?
MCNP (Monte Carlo N-Particle) is a general-purpose Monte Carlo radiation transport code developed by Los Alamos National Laboratory. It is used to simulate the interaction of neutrons, photons, electrons, and other particles with matter. Its common uses include nuclear reactor design, radiation shielding, medical physics, and nuclear criticality safety analysis.
What other technologies are related to MCNP?
MCNP Competitor Technologies
Serpent
Serpent is a Monte Carlo particle transport code, like MCNP, and therefore a direct competitor.
TRIPOLI
No summary available
TRIPOLI is another Monte Carlo particle transport code, similar to MCNP, making it a competitor.
GEANT4
GEANT4 is a Monte Carlo simulation toolkit used in various fields including high energy physics, and is another particle transport code that can perform similar simulations to MCNP.
OpenMC
OpenMC is a Monte Carlo particle transport code focusing on reactor physics calculations, directly competing with MCNP for some applications.
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MONK
No summary available
MONK is a Monte Carlo criticality code, making it a direct competitor for criticality calculations.
FLUKA
No summary available
FLUKA is a Monte Carlo particle transport code used for shielding and dosimetry calculations, competing with MCNP.
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Attila
No summary available
Attila is a discrete ordinates radiation transport code, offering an alternative approach to MCNP's Monte Carlo method, but for similar applications.
GEANT
No summary available
Older version of Geant4, which is a Monte Carlo simulation toolkit used in various fields and is another particle transport code that can perform similar simulations to MCNP.
MCNP Complementary Technologies
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MicroShield
MicroShield is a point kernel code used for shielding calculations, offering a different approach than MCNP, but for similar applications.
SCALE
SCALE includes radiation transport capabilities and criticality safety analysis tools. While parts compete, SCALE provides comprehensive nuclear safety analysis capabilities and nuclear data libraries and preprocessing tools which can complement MCNP.
ORIGEN
No summary available
ORIGEN is a code for calculating the buildup, decay, and processing of radioactive materials, which can provide source terms for MCNP simulations.
Number of organizations that mention technology
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Which job functions commonly mention MCNP?
Nuclear Engineer
10.5% of all Nuclear Engineer jobs mention MCNP
View 240
jobs on Sumble
Physicist
2.6% of all Physicist jobs mention MCNP
View 387
jobs on Sumble
Safety Engineer
0.4% of all Safety Engineer jobs mention MCNP
View 111
jobs on Sumble
Scientist
0.1% of all Scientist jobs mention MCNP
View 462
jobs on Sumble
Radiologist
0.1% of all Radiologist jobs mention MCNP
View 155
jobs on Sumble
Which organizations are mentioning MCNP?
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