Papers using OpenMC

Announcements
21

Verification of the current coupling collision probability method with orthogonal flux expansion for the assembly calculations

22

Neutronic modeling of megawatt-class heat pipe reactors

1 Like
23

The windowed multipole formalism and applications to uncertainty quantification

24

Benchmarking and verification of the OpenMC code for accelerator-based neutron source analyses

25

This one is particularly exciting for shielding applications
Development and benchmarking of the Weight Window Mesh function for OpenMC

Related PR

2 Likes
26

Studies on calculation models of ASTRA critical facility benchmark using OpenMC

27

Congrats to @fsabab on his recent paper:
Modeling and neutronic analysis of pin-cell comprising nuclear fuel with different chemical composition and neutron moderator using Monte Carlo code OpenMC

3 Likes
28

Generation method and verification of pebble type VHTR multigroup cross sections based on OpenMC

29

Comparative Study on Fuel Assembly of Modular Gas-cooled Fast Reactor using MCNP and OpenMC Code

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Fuel Assembly Design Study for Modular Gas Cooled Fast Reactor using Monte Carlo Parallelization Method

31

An Analytic Benchmark for Neutron Boltzmann Transport with Downscattering—Part II: Flux and Eigenvalue Sensitivities to Nuclear Cross Sections and Resonance Parameters

32

Neutronic Design for Heat Pipe Reactor With Annular and Accident Tolerant Fuels

1 Like
33

We wrote a new paper where we implemented solid and annular rod in the same assembly and analyzed that using both OpenMC and DRAGON. The paper has been published in Nuclear Engineering and Design, Elsevier. In case, If anyone is interested,
Assessment of the burnup characteristics of UO2 and MOX fuel in the mixed solid and annular rod configuration - ScienceDirect

3 Likes
34

Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO₂, MOX and (Th/U)O₂ using OpenMC

35

Determination of lattice physics properties and uncertainties in a solid fuel molten salt cooled assembly using OpenMC

1 Like
36

Core depletion analysis of long-life CANDLE gas-cooled fast reactor using OpenMC code

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Arbitrarily Large Neutron Amplification in Subcritical Nuclear Reactors

38

Nice paper by @makeclean:
Uncertainty Propagation in SINBAD Fusion Benchmarks with Total Monte Carlo and Imprecise Probabilities

39

Implementation and Benchmarking of an Automatic Global Variance Reduction Method on OpenMC

1 Like
40

Solving variable-coefficient burnup equations by exponential Rosenbrock methods