Papers using OpenMC

Announcements
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I thought I’d start a topic for anyone wanting to share a paper they wrote in which OpenMC was used, or if you came across a paper that used OpenMC.

Here’s one that I just came across using OpenMC for some studies on a lead-cooled reactor:
https://www.researchgate.net/publication/344376283_Investigation_of_a_Self-actuated_Gravity-driven_Shutdown_System_in_a_Small_Lead-cooled_Reactor

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Here’s another one focused on the study of tritium breeding ratios in a fusion reactor:
Neutronic comparison of liquid breeders for ARC-like reactor blankets

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SARAX: A new code for fast reactor analysis part I: Methods

On the equivalence of reaction rate in energy collapsing of fast reactor code SARAX

Investigation of the VVER-1000 reactor pressure vessel neutron fluence and displacement per atom using MCNP6

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A paper by @pshriwise:
Implementation and verification of PyNE R2S with DAG-OpenMC

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The PyNE-Based Burnup Analysis Method for Accelerator-Driven Subcritical Systems

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Conservation of migration area by transport cross sections using Cumulative Migration Method in deterministic heterogeneous reactor transport analysis

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Core and blanket thermal–hydraulic analysis of a molten salt fast reactor based on coupling of OpenMC and OpenFOAM

Neutronics analysis of the stellarator-type fusion-fission hybrid reactor based on the CAD optimization method

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ANS 2020 Annual Meeting Summary
DAG-OpenMC: CAD-Based Geometry in OpenMC

Presentation Slides

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Dear all,
I wrote a paper for this PHYSOR 2020, called: " Verification of the OpenMC homogenized MYRRHA-1.6 core model" https://publications.sckcen.be/portal/en/publications/verification-of-the-openmc-homogenized-myrrha16-core-model(422d637f-2685-479d-bb3e-217de2d7ad5d)/export.html
Is about XS homogenization for the research reactor known as MYRRHA. Thank you.

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A recent paper by @Shimwell, @makeclean, and company:
High-speed generation of neutronics-ready CAD models for DEMO design

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Reflector materials selection for core design of modular gas‐cooled fast reactor using OpenMC code

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Extension and benchmarking of the OpenMC code for accelerator-based neutron source applications

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Nice paper using OpenMC to look into thermal scattering for U-Mo alloys:
Crystal binding effects on neutron scattering and criticality in U–Mo fuels

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Extended development of a Monte Carlo code OpenMC for fuel cycle simulation of molten salt reactor

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An interesting paper on a method for assessing source convergence that doesn’t rely on a mesh as our current Shannon entropy implementation does:
An effective mesh-free fission source convergence indicator for Monte Carlo k-Eigenvalue problems

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A study of criticality and thermal loading in a conceptual micronuclear heat pipe reactor for space applications

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A few recent papers:

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Criticality Analysis of HTR-10 using an Open Source Monte Carlo code OpenMC

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Low temperature effects on PWR fuel assembly criticality calculations