| No. | Year | Title | Venue |
|---|---|---|---|
| 01 | 2014 | Calculation of the Radial and Axial Flux and Power Distribution for a CANDU 6 Reactor with both the MCNP6 and Serpent CodesM. S. Hussein, H. W. Bonin, B. J. LewisBuilt full 3-D MCNP6 and Serpent models of the CANDU 6 (Gentilly-2) core with 37-element bundles and compared flux and power-density maps; Serpent matched MCNP6 and was the more efficient of the two. | PBNC 2014Conference · Vancouver |
| 02 | 2014 | Burnup Calculation of a CANDU6 Reactor Using the Serpent and MCNP6 CodesM. S. Hussein, H. W. Bonin, B. J. LewisValidated whole-core burnup in MCNP6 and Serpent against WIMS-AECL, the industry lattice code, with k-effective versus burnup in excellent agreement. Still cited in 2025 work on SMR spent fuel in CANDU. | PBNC 2014Conference · Vancouver |
| 03 | 2015 | Design of a MultiSpectrum CANDU-based Reactor, MSCR, with 37-Element Fuel Bundles Using Serpent CodeM. S. Hussein, H. W. Bonin, B. J. Lewis, P. K. ChanFirst public design of the Multispectrum CANDU Reactor: an inner helium-cooled fast core inside a CANDU 6 thermal core, both with 37-element bundles, sized for criticality safety and studied across several U-235 enrichments. | ICMSNSE 2015Conference · Ottawa |
| 04 | 2017 | Design of a Multispectrum CANDU Reactor for Burning Actinides: An Approach of Non-Proliferation and Nuclear Fuel RecyclingMohamed Salah HusseinSix full-core MSCR models in Serpent. The best uranium model destroyed 10.7 % of fissile isotopes in 1000 days; the best plutonium model 15.3 % over 2600 days. Either core alone is subcritical, so disturbing one shuts the reactor down. | PhD thesis, RMCThesis · Kingston, Ontario |
§ Expertise Monte Carlo reactor and radiation analysis
Full-core Monte Carlo models, built to scale.
Three-dimensional reactor and radiation-transport models in MCNP and Serpent, built at full-core scale on parallel computing clusters.
Full-core models
Monte Carlo codes track individual neutrons and photons through an exact three-dimensional geometry instead of averaging it away, which is why they serve as the reference when other methods need checking. His models are full-core, not single cells. At the Royal Military College of Canada he built complete CANDU 6 cores for the Gentilly-2 design in both MCNP6.1 and Serpent, with all 380 fuel channels and 37-element bundles, and computed the multiplication factor, radial and axial flux, and power-density distributions from them. The two codes agreed closely, and Serpent was the more efficient in computing time.
The same models carried burnup. He normalised track-length flux to reactor power and followed k-effective against burnup through the core’s life, then compared the results with WIMS-AECL, the industry lattice code. He also defined a channel-flux-to-fuel-flux ratio for turning tallies into per-bundle power.
New reactor designs
His doctoral thesis used these methods to design a reactor that did not exist: the Multispectrum CANDU Reactor, a helium-cooled fast-neutron core inside a CANDU thermal core, intended to burn surplus weapons-grade uranium and plutonium. He built six full-core models in Serpent, varying core radius, lattice pitch, fuel and the number of fast-core channels, and tracked excess reactivity, flux and power shape, burnup and actinide inventories for each. The best uranium model destroyed 10.7 % of its fissile isotopes over 1000 days without refuelling; the best plutonium model destroyed 15.3 % over 2600 days.
The thesis runs to 363 pages, with 397 figures. Calculations at this scale ran on parallel clusters at the High Performance Computing Virtual Laboratory (HPCVL), Queen’s University.
Radiation transport and shielding
The same codes solve radiation-transport problems outside the core. From 2017 to 2020 he taught MCNP and Serpent for radiation transport, core modelling, shielding optimisation and radiological dose assessment, and his master’s degree in engineering management included shielding engineering and advanced microdosimetry. In 2016 he attended the University of Michigan’s MCNPX-PoliMi workshop, and in 2026 he completed the OECD Nuclear Energy Agency’s course on OpenMC, the open-source code now common in small-modular-reactor work.
§1 Papers
The work this rests on.
§2 Related
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§3 Contact
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