Cumulative migration method for computing multi-group transport cross sections and diffusion coefficients with Monte Carlo calculations

In nuclear reactor physics analysis, fast accurate deterministic methods are needed for the many full-core calculations required for safe and efficient operation of nuclear power plants. Multi-group diffusion coefficients and transport cross sections are the crucial parameters that balance efficiency and accuracy in full-core simulations. However, it is not clear what definition of diffusion coefficients and transport cross sections should be employed or what "transport properties" are preserved by the numerous approximations available in the literature. Among the sources of error associated with efficient deterministic simulations of nuclear reactors, whether diffusion or transport theory, the anisotropy of neutron scattering introduces one major challenge for achieving highly accurate eigenvalues and power distributions. Anisotropic scattering has a significant impact on the neutron spatial migration, which is an important transport property in nuclear reactor systems.

Name thesis:degree_name

Doctoral

Department dc:contributor.department

Massachusetts Institute of Technology. Department of Nuclear Science and Engineering

Grantor dc:publisher

Massachusetts Institute of Technology

Year dc:date.issued

2020

Author dc:creator

• Liu, Zhaoyuan,Ph. D.Massachusetts Institute of Technology.

Advisor dc:contributor.advisor

• Kord S. Smith and Benoit Forget.

• Nuclear Science and Engineering.

Statement dc:rights

• MIT theses may be protected by copyright. Please reuse MIT thesis content according to the MIT Libraries Permissions Policy, which is available through the URL provided.

Licence dc:rights.uri

• http://dspace.mit.edu/handle/1721.1/7582

Language dc:language.iso

eng

Handle dc:identifier.uri

https://hdl.handle.net/1721.1/129924

OAI identifier oai:identifier

oai:dspace.mit.edu:1721.1/129924