Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices

Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computational perspective. Full-core neutronic calculations for CANDU reactors are therefore performed customarily using two-energy-group diffusion theory (no angular dependence) for a node-homogenized reactor model. The work presented here is concerned with reducing the loss in accuracy entailed when going from Transport to Diffusion. To this end a new method of calculating the diffusion coefficient was developed, based on equating the neutron balance equation expressed by the transport equation with the neutron balance equation expressed by the diffusion equation. The technique is tested on a simple twelve-node model and is shown to produce transport-like accuracy without the associated computational effort.

Name thesis:degree_name

Master of Science (MSc)

Discipline thesis:degree_discipline

Modelling and Computational Science

Grantor

University of Ontario Institute of Technology

Year dc:date.issued

2010

Author dc:creator

• Patel, Amin

Advisor dc:contributor.advisor

• Nichita, Eleodor M.

• Applied reactor physics
• Transport theory
• Diffusion theory
• CANDU
• Nuclear reactor

Language dc:language.iso

en

Handle dc:identifier.uri

https://hdl.handle.net/10155/87

OAI identifier oai:identifier

oai:ontariotechu.scholaris.ca:10155/87