| No. | Year | Title | Venue |
|---|---|---|---|
| 01 | 2013 | Numerical Verification of the Theory of Coupled Reactors for a Deuterium Critical Assembly Using MCNP5M. S. Hussein, H. W. Bonin, B. J. LewisChecked multipoint coupled-reactor theory against MCNP5 on the two-region Deuterium Critical Assembly: multiplication factors and coupling coefficients agreed as the water levels in each core were varied. | CNS Annual 2013Conference · Toronto |
| 02 | 2013 | Numerical Verification/Validation of the Theory of Coupled Reactors for Deuterium Critical Assembly, using MCNP5 and Serpent CodesM. S. Hussein, B. J. Lewis, H. W. BoninRepeated the coupled-reactor verification with both MCNP5 and Serpent and found excellent agreement, establishing Serpent as valid for multipoint coupled-core calculations. Also given as a talk. | CANDU Fuel 2013Conference · Kingston, Ontario |
| 03 | 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 |
| 04 | 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 |
| 05 | 2016 | The application of UWB1 nuclear fuel depletion code on a CANDU fuel bundleM. Lovecký, R. Škoda, M. S. Hussein, J. J. Song, P. K. ChanBenchmarked the University of West Bohemia's fast depletion code UWB1 on the 37-element CANDU bundle against WIMS-AECL, Serpent and MCNP6, so burnable-absorber studies can run in seconds per depletion step instead of hours. | Prog. Nucl. EnergyJournal |
| 06 | 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 |
MCNP, used to check reactor models.
His published MCNP work, from 2013, models coupled reactor cores and a full CANDU 6 core, cross-checked against Serpent and, for burnup, against the industry code WIMS-AECL. He has also taught the code.
What MCNP is
MCNP (Monte Carlo N-Particle) is a general-purpose radiation transport code developed at Los Alamos National Laboratory. It follows individual neutrons, photons and other particles through a three-dimensional model of a reactor, a shield or a detector, one history at a time, using continuous-energy nuclear data. Averaged over many millions of histories, the results give the multiplication factor of a core, the flux and power in each region, and dose behind a shield. Because it models geometry and physics with few simplifications, it is widely used as a reference against which faster design codes are checked.
How he has used it
His first use was theoretical. In 2013 he built MCNP5 models of the Deuterium Critical Assembly, a two-region experimental assembly, to test multipoint coupled-reactor theory: multiplication factors and coupling coefficients from the theory agreed with MCNP5 as the water levels in each core were changed. He then repeated the study with both MCNP5 and Serpent.
In 2014 he moved to a power reactor. He built a full three-dimensional MCNP6.1 model of the CANDU 6 core at Gentilly-2, down to the 37-element fuel bundles, and computed radial and axial flux and power-density maps, turning track-length flux tallies into power per bundle. A second paper followed the same core through burnup and compared k-effective against WIMS-AECL, the lattice code the Canadian industry uses; the agreement was excellent. This validation became a chapter of his doctoral thesis, and MCNP6 was also one of the reference codes when the fast depletion code UWB1 was benchmarked on the CANDU fuel bundle.
Full-core runs of this size need many histories, so he ran them in parallel on the HPCVL high-performance computing clusters at Queen’s University.
Shielding, dose and teaching
MCNP is as much a shielding and dosimetry code as a reactor code. In 2016 he completed the MCNPX-PoliMi training workshop at the University of Michigan, on the variant of the code built for detector and measurement simulation. From 2017 to 2020 he taught Radiation Transport and Computational Modelling, which covered MCNP and Serpent for radiation transport, core modelling, shielding optimisation and dose assessment.
Why two codes
After the first coupled-core paper he ran MCNP alongside a second code rather than on its own. A model that two independently written codes agree on is far easier to trust than one run in either code alone. That habit is the thread through his published work. In June 2026 he completed the OECD Nuclear Energy Agency’s Introduction to OpenMC.
§1 Papers
The work this rests on.
§2 Related