| No. | Year | Title | Venue |
|---|---|---|---|
| 01 | 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 |
| 02 | 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 |
| 03 | 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 |
| 04 | 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 |
| 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 |
Serpent, from verification to a new reactor design.
He verified the Serpent Monte Carlo code against MCNP and WIMS-AECL on coupled cores and a full CANDU 6 core, then used it to build all six models of the reactor in his doctoral thesis.
What Serpent is
Serpent is a continuous-energy Monte Carlo code for reactor physics and burnup, developed at VTT Technical Research Centre of Finland. Like MCNP, it follows neutrons one at a time through a three-dimensional model. Unlike a general-purpose transport code, it was written with reactor work in mind: it has fuel depletion built in, so a single model can track how the fuel changes over months of operation, and its way of tracking particles makes large lattice geometries such as a full reactor core fast to run.
Verifying it before relying on it
He did not start by trusting it. In 2013 he repeated a verification of coupled-reactor theory on the Deuterium Critical Assembly, a two-region experimental assembly, with both MCNP5 and Serpent. The two codes agreed closely, which established Serpent as valid for calculations where two cores are neutronically coupled, the situation his later reactor design depends on.
In 2014 he built the CANDU 6 core at Gentilly-2 in Serpent 1.1.19 and in MCNP6.1, with every 37-element fuel bundle modelled, and compared the multiplication factor and the radial and axial flux and power distributions. They agreed, and Serpent was the more efficient of the two in computing time. A companion paper followed the core through burnup and compared k-effective against WIMS-AECL, the lattice code used by the Canadian industry, again with excellent agreement. Serpent was also one of the reference codes when the fast depletion code UWB1 was benchmarked on the CANDU fuel bundle.
The Multispectrum CANDU Reactor
With Serpent verified for coupled cores and validated on CANDU 6, he used it for the design work in his doctoral thesis. The Multispectrum CANDU Reactor puts a helium-cooled fast-neutron core inside a CANDU thermal core to burn surplus weapons-grade uranium or plutonium. He built six full-core models in Serpent, varying the fast core’s radius, lattice pitch, fuel and number of channels, and followed each through burnup: excess reactivity, flux and power, and the inventory of fissile isotopes over time. The best uranium model destroyed 10.7 % of its fissile isotopes in 1000 days; the best plutonium model destroyed 15.3 % over 2600 days.
That is the pattern of all his Serpent work: check the code against another code and against the industry reference first, then use it for something new.
§1 Papers
The work this rests on.
§2 Related