Calculation of Neutron Flux in SPRR-300 Reactor Irradiation Channel with MCNP Code

DOU Hai-feng; DAI Jun-long

Feb. 2006

Article Contents

Article Navigation > Nuclear Power Engineering > 2006 > 27(1): 51-54

DOU Hai-feng, DAI Jun-long. Calculation of Neutron Flux in SPRR-300 Reactor Irradiation Channel with MCNP Code[J]. Nuclear Power Engineering, 2006, 27(1): 51-54.

Citation:

DOU Hai-feng, DAI Jun-long. Calculation of Neutron Flux in SPRR-300 Reactor Irradiation Channel with MCNP Code[J]. Nuclear Power Engineering, 2006, 27(1): 51-54.

DOU Hai-feng, DAI Jun-long. Calculation of Neutron Flux in SPRR-300 Reactor Irradiation Channel with MCNP Code[J]. Nuclear Power Engineering, 2006, 27(1): 51-54.

Citation:

DOU Hai-feng, DAI Jun-long. Calculation of Neutron Flux in SPRR-300 Reactor Irradiation Channel with MCNP Code[J]. Nuclear Power Engineering, 2006, 27(1): 51-54.

PDF( 193 KB)

Calculation of Neutron Flux in SPRR-300 Reactor Irradiation Channel with MCNP Code

Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang, 621900, China

Received Date: 2004-09-27
Rev Recd Date: 2005-01-26

Available Online: 2025-07-29

Abstract

Abstract

The issue predigests geometric structure of SPRR-300 reactor core with repeated structure geometry feature of MCNP code and sets up the irradiation channel mathematical model to calculate the neutron flux.The calculated results accorded well with the experimental data shows that the mathematical model in this issue is reasonable and feasible.
- SPRR-300 reactor core,
- Irradiation channel,
- MCNP code,
- Neutron flux,
- Mathematical model