Capabilities overview of the MORET 5 Monte Carlo code

被引：24
作者：
Cochet, B. [1 ]
Jinaphanh, A. [1 ]
Heulers, L. [1 ]
Jacquet, O.
机构：
[1] IRSN, LNC, SNC, PSN EXP, F-92262 Fontenay Aux Roses, France
来源：
ANNALS OF NUCLEAR ENERGY | 2015年 / 82卷
关键词：
MORET; Monte Carlo; Neutron transport; Criticality; Reactor modeling; Validation;
D O I：
10.1016/j.anucene.2014.08.022
中图分类号：
TL [原子能技术]; O571 [原子核物理学];
学科分类号：
0827 ; 082701 ;
摘要：
The MORET code is a simulation tool that solves the transport equation for neutrons using the Monte Carlo method. It allows users to model complex three-dimensional geometrical configurations, describe the materials, define their own tallies in order to analyse the results. The MORET code has been initially designed to perform calculations for criticality safety assessments. New features has been introduced in the MORET 5 code to expand its use for reactor applications. This paper presents an overview of the MORET 5 code capabilities, going through the description of materials, the geometry modeling, the transport simulation and the definition of the outputs. (C) 2014 Elsevier Ltd. All rights reserved.
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收藏
页码：74 / 84
页数：11
相关论文
共 27 条
[1]  
Akhiezer N. I., 1965, CLASSICAL MOMENT PRO
[2]  
Anno J., 2003, INT C NUCL CRIT TOK
[3]  
Cochet B., 2013, JOINT INT C SUP NUCL
[4]   AN ANALYTIC APPROXIMATION TO THE DISTRIBUTION OF LILLIEFORS TEST STATISTIC FOR NORMALITY [J].
DALLAL, GE ;
WILKINSON, L .
AMERICAN STATISTICIAN, 1986, 40 (04) :294-296
[5]  
Forestier B., 2008, PHYSOR INT SWITZ
[6]  
Gomit J.M., 2011, INT C NUCL CRIT 2011
[7]  
Hendricks J.S., 2005, LAUR058225 LOS AL NA
[8]  
Heulers L, 2011, INT C NUCL CRIT 2011
[9]  
Hurwitz H, 1964, NAVAL REACTOR PHYS H
[10]  
Jinaphanh A., 2012, PHYSOR KNOXV TENN US
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