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MCNP

Known as: MCNPX, Monte Carlo N-Particle Transport 
Monte Carlo N-Particle Transport Code (MCNP) is a software package for simulating nuclear processes. It is developed by Los Alamos National… 
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Related topics

Related topics

8 relations
Computational physicsConstructive solid geometryFLUKA Particle Transport CodeGeant4

Papers overview

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2016
2016
A quantitative PGNAA study for use in aqueous solution measurements using Am–Be neutron source and BGO scintillation detector
2010
2010
Neutron production in a Pb/U-setup irradiated with 0.7–2.5 GeV protons and deuterons
2010
2010
Digitized neutron imaging with high spatial resolution at a low power research reactor: Applications to steel and rock samples
Highly Cited
2008
Highly Cited
2008
A Functional Monte Carlo Method for k-Eigenvalue Problems
Abstract In Monte Carlo simulations of k-eigenvalue problems for optically thick fissile systems with a high dominance ratio, the… 
2005
2005
Evaluation of the activity of irradiated graphite in the Ignalina Nuclear Power Plant RBMK-1500 reactor
Ignalina NPP Unit 1 with the RBMK-1500 type reactor enters the preparatory stage for its decommissioning. A big volume of… 
2005
2005
Source efficiency and high-energy neutronics in accelerator-driven systems
Transmutation of plutonium and minor actinides in accelerator-driven systems (ADS) is being envisaged for the purpose of reducing… 
2005
2005
Immobilization of metal-containing nanoparticles on the surface of polytetrafluoroethylene nanogranules
2004
2004
14 MeV neutron-induced SEU in SRAM devices
In this paper we explore the sensitivity of commercial static random-access memory devices to 14 MeV neutrons. The effects on… 
Highly Cited
2003
Highly Cited
2003
Design strategies for optimizing high burnup fuel in pressurized water reactors
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2003. 
2002
2002
The biological shield of a high-intensity spallation source: a Monte Carlo design study
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