| 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 | 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 |
§ Research Monte Carlo verification and validation
Verification first, design second.
Six of the seven papers he published between 2013 and 2016 check a code against another code, or against theory, before any design result is used.
The question
A Monte Carlo code follows simulated neutrons one at a time through a model of a reactor: each flight, collision, absorption and fission is sampled from nuclear data. Run enough histories and the averages give the multiplication factor, the flux and the power in every region. The method makes few approximations, which is why it is used to check faster, simplified codes. But a Monte Carlo result is only as good as the geometry, the materials, the nuclear data and the tallies the analyst sets up. A second, independent model is the most direct check that none of those went wrong.
That check is the thread through his doctoral work at the Royal Military College of Canada. Before the Multispectrum CANDU Reactor could be designed in Serpent, he had to show that Serpent and MCNP gave the same answers on problems where the answer could be tested.
What he did
Coupled-core theory on the Deuterium Critical Assembly (2013). The reactor concept in his thesis depends on the theory of two neutronically coupled cores. He modelled the two-region Deuterium Critical Assembly in MCNP5 and computed multiplication factors and coupling coefficients as the water levels in the inner and outer cores changed. He then repeated the calculation in Serpent. The two codes and the theory agreed, which established Serpent as valid for multipoint coupled-core calculations. This is thesis chapter 5, and two conference papers.
Full-core CANDU 6 in two codes (2014). He built full three-dimensional models of the Gentilly-2 CANDU 6 core, 380 fuel channels of 37-element bundles, in both MCNP6 and Serpent, and compared the radial and axial flux and power-density distributions. They agreed, and Serpent was the more efficient of the two in computing time.
Burnup against the industry code (2014). He then ran whole-core burnup in both codes and compared k-effective against burnup with WIMS-AECL 3.1.2.1, the lattice code used by the Canadian industry. The agreement was excellent. That benchmark is still cited: a 2025 paper in Nuclear Technology on using SMR spent fuel in CANDU 6 cites it.
Why it matters beyond the thesis
The habit of building the same problem twice, in codes with different nuclear-data handling and tally machinery, and explaining every difference, is the skill his papers document. In June 2026 he completed the OECD Nuclear Energy Agency’s Introduction to OpenMC.
§1 Papers
The work this rests on.
§2 Related