Results 121 to 130 of about 16,330 (259)

MCNP analysis of the FOEHN critical experiment

, 1993
A very high fidelity MCNP model of the Franco-German FOEHN critical experiment has been developed. The results obtained show a high degree of agreement with each of the three configurations of the experiment.
Wemple, C. A., Rubio, G. A., Ougouag, A. M., Ryskamp, J. M.   +3 more
core   +1 more source

New developments enhancing MCNP for criticality safety

, 1993
Since the early 80`s MCNP has had three estimates of k{sub eff}: collision, absorption, and track length. MCNP has also had collision and absorption estimators of removal lifetime.
McKinney, G. W., Forster, R. A., Hendricks, J. S.   +2 more
core  

MCNP® Code Version 6.3.0 Build Guide

, 2022
J. Bull, C. Josey, J. Kulesza, M. Rising
semanticscholar   +1 more source

MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis [PDF]

, 2005
The UNIX BASH (Bourne Again Shell) script MCWO has been developed at the Idaho National Engineering and Environment Laboratory (INEEL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN2.
Chang, Gray S
core  

Status of electron transport in MCNP{trademark}

, 1995
In recent years, an ongoing project within the radiation transport group (XTM) at Los Alamos National Laboratory has been the implementation and validation of an electron transport capability in the Monte Carlo code NICNP.
Hughes, H.G.
core  

MCNP code [PDF]

, 1984
The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain.
Cramer, S.N.
core  

MCNP User Manual : Code Version 6.2

, 2018
Buku ini berisi manual penggunaan dan teori MCNP terkini dan historis.x, 732 ...
Werner, Christopher J
core  

MCNP application for the 21 century

, 2000
The Los Alamos National Laboratory (LANL) Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron, photon, and electron radiation transport applications.
McKinney, M.C.
core  

Synthesis and evaluating gamma shielding features of YBa₂Cu₃O₇(Bi₂O₃) ceramic composite using Monte Carlo method via GEANT4 and MCNP codes. [PDF]

Sci Rep
Soltani E, Saray AA.
europepmc   +1 more source

MCODE, Version 2.2: An MCNP-ORIGEN DEpletion Program

, 2008
MCODE Version 2.2 is a linkage program, which combines the continuous-energy Monte Carlo code, MCNP-4C, and the one-group depletion code, ORIGEN2, to perform burnup calculations for nuclear fission reactor systems. MCNP is used as the advanced physics
Hejzlar, Pavel, Xu, Z.
core