ARRC: A random ray neutron transport code for nuclear reactor simulation

The Serpent Monte Carlo code: Status, development and applications in 2013

Jaakko Leppänen, Maria Pusa, Tuomas Viitanen et al. · Annals of Nuclear Energy · 2014 · 1.4K citations

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MCNP-A general Monte Carlo code for neutron and photon transport

J.F. Briesmeister · CERN Document Server (European Organization for Nuclear Research) · 1986 · 709 citations

Engineering, Nuclear Physics, Physics +9

Nuclear Reactor Analysis

Noel Corngold · Nuclear Science and Engineering · 1977 · 543 citations

Nuclear Reactor Analysis, Engineering, Nuclear Physics +14

The OpenMC Monte Carlo particle transport code

Paul Romano, Benoit Forget · Annals of Nuclear Energy · 2012 · 446 citations · Full text

Engineering, Physics, Natural Sciences +6

Knights Landing: Second-Generation Intel Xeon Phi Product

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Hardware Security, System On Chip, Manycore Processor +14