Published July 2006 | Version v1
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Utilization of MCNP code in the research and design for China advanced research reactor

Creators

  • 1. Dept. of Reactor Engineering Research and Design, China Inst. of Atomic Energy, Beijing (China)

Description

MCNP, which is the internationalized neutronics code, is used for nuclear research and design in China Advanced Research Reactor (CARR). MCNP is an important neutronics code in the research and design for CARR since many calculation tasks could be undertaken by it. Many nuclear parameters on reactor core, the design and optimization research for many reactor utilizations, much verification for other nuclear calculation code and so on are conducted with help of MCNP. (author)

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Part of:
Proceedings of the FNCA 2004 workshop on the utilization of research reactors (Contract research)

Additional details

Publishing Information

Imprint Title
Proceedings of the FNCA 2004 workshop on the utilization of research reactors (Contract research)
Imprint Pagination
165 p.
Journal Page Range
p. 51-62
Report number
JAEA-Conf--2006-005

Conference

Title
FNCA 2004 workshop on the utilization of research reactors
Dates
13-21 Jan 2005
Place
Bangkok (Thailand)

Optional Information

Notes
3 refs., 6 figs., 4 tabs.