37th Annual CNS Conference - 2017 June 04-07

Presented at:
37th Annual CNS Conference
2017 June 04-07
Niagara Falls
Session Title:
1A5 - Reactor and Radiation Physics (I)

P. Schwanke (UOIT)
E. Nichita (UOIT)


Verification of neutronic-analysis codes involves comparing their results to some benchmark, whose correctness is accepted. Analytical results are particularly useful as benchmarks because they are exact. This work presents analytic solutions of the two-group neutron-diffusion equation for three two-region reactor configurations comprised of fuel and reflector, which are symmetric about the origin: an infinite slab, an infinite cylinder and a sphere. The solutions are based on examples appearing in advanced reactor-physics texts, which have been extended to account for all the terms appearing in the two-group diffusion equation (e.g. up-scattering) and to include a method for determining keff for a given set of material properties and reactor dimensions. Results are generated for macroscopic cross-sections representative of PHWR configurations, namely 37-element zero-burnup natural-uranium fuel bundles with D2O coolant and moderator bounded by a D2O reflector. These results offer a benchmark for existing and future two-group diffusion codes used for PHWR configurations.

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