Neutron transport and activation comparison between OpenMC and FISPACT-II in ARC-class reactor [PDF]
In a fusion reactor, high-energy neutron fluxes strike the materials causing radiation damage and triggering nuclear reactions that alter the chemical composition of the materials through transmutation.
Raffaella Testoni +3 more
core +1 more source
OpenMC Interpretation of FNS SINBAD Shielding Benchmark Experiments
The Fusion Neutron Source (FNS) clean benchmark experiments on tungsten, vanadium, and beryllium assemblies from the SINBAD (Shielding Integral Benchmark Archive and Database) are analyzed to experimentally validate OpenMC (version 0.14.1-dev) fusion ...
Segantin, Stefano +2 more
core +1 more source
Evaluating machine learned nuclear data precision in full core nuclear reactor Monte Carlo neutronics and computational efficiency analyses. [PDF]
Sci RepHashemi A, Macián-Juan R, Ohlerich M.
europepmc +1 more source
Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list. [PDF]
PLoS One, 2018 Merk B, Litskevich D, Gregg R, Mount AR.
europepmc +1 more source
Sensitivity-informed framework for enrichment distribution in MNR for thermal performance enhancement. [PDF]
Sci RepAziz U +5 more
europepmc +1 more source
Development of a novel machine learning-based adaptive resampling algorithm for nuclear data processing. [PDF]
Sci RepHashemi A, Macián-Juan R, Ohlerich M.
europepmc +1 more source
ZED-2 benchmarks performed in OpenMC and Serpent2: A validation exercise for OpenMC applications
Annals of Nuclear EnergyPeter J. Kriemadis, Adriaan Buijs
openaire +1 more source
Prediction and assessment of optimal concrete compositions for overall radiation protection and reduced global warming potential. [PDF]
Sci RepSaxena S, Sharma H.
europepmc +1 more source
Neutronics design of shutdown and control systems for a Zero Power Experiments of chloride-based molten salt fast reactor. [PDF]
PLoS OneJain L +5 more
europepmc +1 more source
RTP TRIGA Reactor Kinetics Parameters from OpenMC
The current work uses Iterated Fission Probability (IFP) routine that was recently implemented in OpenMC to calculate reactor kinetics parameters. IFP is calculated from the product of the multiplication factors tracked across the L+1 generations of fission progenies.
openaire +1 more source