Modeling and Simulation in Nuclear Energy

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Modeling and simulation play a crucial role in engineering, particularly in nuclear energy, where finite element methods are applied to assess nuclear plants and reactor performance under various conditions. The American Nuclear Society's Nuclear Technology journal features articles on codes developed by the US DOE Idaho National Laboratory, highlighting advancements in computational frameworks like MOOSE. This open-source Multiphysics Object-Oriented Simulation Environment facilitates nuclear power applications but has restricted access to certain modules. Additionally, OpenMC has made significant strides in performing burnup calculations, showcasing the evolving capabilities in the field. Overall, these developments underscore the importance of computational tools in enhancing nuclear engineering practices.
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Modeling and simulation, or computational physics/chemistry, is a large and important part of engineering. In nuclear energy, there are applications of finite element methods (and occasionally finite different or finite volume depending on the problem) applied to nuclear plants, nuclear reactors and nuclear fuel performance/behavior under normal, off-normal, and abnormal/transient conditions.

ANS's Nuclear Technology has a nice set of articles on some of the codes developed by US DOE Idaho National Laboratory (INL). Similar codes have been developed by other institutions and in other nations. OECD/NEA also provides a set of codes.

Nuclear Technology

Volume 207, 2021 - Issue 7: Special issue on the MOOSE Multiphysics Computational Framework

The first article introduces the MOOSE Multiphysics Computational Framework for Nuclear Power Applications: A Special Issue of Nuclear Technology
https://www.tandfonline.com/doi/full/10.1080/00295450.2021.1915487?src=recsys

It is currently open source.

Keyword: Multiphysics Object-Oriented Simulation Environment (MOOSE) framework
 
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This is a cool thing, and the MOOSE framework is indeed open source and can be found here,
https://github.com/idaholab/moose

Sadly the nuclear engineering modules, even Sockeye that just does heat pipe modeling, are behind layers of your organisation must be registered with our organisation and we will vet your application individually and even then we probably won't give you source code access.

Meanwhile OpenMC does burnup calculations now and that is absolutely blowing my mind.
 
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Hello everyone, I am currently working on a burnup calculation for a fuel assembly with repeated geometric structures using MCNP6. I have defined two materials (Material 1 and Material 2) which are actually the same material but located in different positions. However, after running the calculation with the BURN card, I am encountering an issue where all burnup information(power fraction(Initial input is 1,but output file is 0), burnup, mass, etc.) for Material 2 is zero, while Material 1...
Hi everyone, I'm a complete beginner with MCNP and trying to learn how to perform burnup calculations. Right now, I'm feeling a bit lost and not sure where to start. I found the OECD-NEA Burnup Credit Calculational Criticality Benchmark (Phase I-B) and was wondering if anyone has worked through this specific benchmark using MCNP6? If so, would you be willing to share your MCNP input file for it? Seeing an actual working example would be incredibly helpful for my learning. I'd be really...
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