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)
Files
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)
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 37122193
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- CHINA; DESIGN; GROUP CONSTANTS; HEAVY WATER; M CODES; MONTE CARLO METHOD; MULTIPLICATION FACTORS; NEUTRON BEAMS; NEUTRON SOURCES; NEUTRON TRANSPORT; POWER DISTRIBUTION; RESEARCH REACTORS; THERMAL COLUMNS; WATER COOLED REACTORS
- Descriptors DEC
- ASIA; BEAMS; CALCULATION METHODS; COMPUTER CODES; CROSS SECTIONS; DEUTERIUM COMPOUNDS; DIMENSIONLESS NUMBERS; HYDROGEN COMPOUNDS; NEUTRAL-PARTICLE TRANSPORT; NUCLEON BEAMS; OXYGEN COMPOUNDS; PARTICLE BEAMS; PARTICLE SOURCES; RADIATION SOURCES; RADIATION TRANSPORT; REACTORS; RESEARCH AND TEST REACTORS; WATER
Optional Information
- Notes
- 3 refs., 6 figs., 4 tabs.