Syed Alam said:I would be grateful if you kindly provide an elaborate explanation of pin-by-pin simulations with and without spatial homoge
My query is completely pertaining to Reactor Physics. I am just wondering are you from Nuclear Reactor Physics background?anorlunda said:That asks too much of us. We don't do your work for you.
Feasibility depends on time and money, so your question is too broad.
Can you rephrase with a more specific question?
Can you summarize your own research on that topic?
Please elaborate on "higher resolution", i.e., to what level of resolution is one referring. In the past, the challenge has been one of computational resources and practicality. For what purpose would one need higher resolution?Syed Alam said:How can we achieve "higher-resolution" 3D pin-by-pin core simulators "without spatial homogenization" for the nuclear reactor core (PWR/BWR) calculations?
Yes.In regards to this question, is it okay to perform "pin-by-pin fine-mesh core calculation" and/or "cell-heterogeneous detailed transport calculation by the method of characteristics (MOC)"?
Massive thanks. I have got a clear sense from your kind reply.rpp said:There are two methods to solve full-core problems without pin homogenization that are currently in-use today.
The first is Monte Carlo methods. Full-core problems are still very big problems to run, but it has been done with
codes like MC21, SHIFT, and OPENMC. You should be able to perform a web search on these terms.
The second method is the so-called "2D/1D" methods where you run a 2D MOC code for each plane in the core, then tie the planes together using 1D diffusion or transport solutions. Some codes that use this method are MPACT, DeCart, and nTracer.
A quick web search shows this paper for MPACT:
http://glc.ans.org/nureth-16/data/papers/13762.pdf
One of the full-core benchmark problems that researchers are currently running is the BEAVRS benchmark from MIT.
https://www.tandfonline.com/doi/full/10.1080/00223131.2015.1038664?src=recsys&
https://www.casl.gov/sites/default/files/docs/CASL-U-2015-0183-000.pdf
This should give you a start.
Dear Astronuc, your response is very helpful providing in-depth explanation of my questions. I truly appreciate your kind and expert answerAstronuc said:Please elaborate on "higher resolution", i.e., to what level of resolution is one referring. In the past, the challenge has been one of computational resources and practicality. For what purpose would one need higher resolution?
Yes.
As rpp indicated, there are various efforts underway to develop more sophisticated methods. Some additional examples for consideration.
Kenichi TADA , Akio YAMAMOTO , Yoshihiro YAMANE & Yasunori
KITAMURAY (2008) Applicability of the Diffusion and Simplified P3 Theories for Pin-by-Pin
Geometry of BWR, Journal of Nuclear Science and Technology, 45:10, 997-1008
https://doi.org/10.1080/18811248.2008.9711885 or https://www.tandfonline.com/doi/pdf/10.1080/18811248.2008.9711885
Mitchell T.H. Young
Orthogonal-Mesh, 3-D Sn with Embedded 2-D Method of
Characteristics for Whole-Core, Pin-Resolved Reactor Analysis
https://deepblue.lib.umich.edu/bitstream/handle/2027.42/135759/youngmit_1.pdf?sequence=1&isAllowed=y
Most of my experience is fuel performance analysis (modeling and simulation) of individual fuel rods for calculations of PCI, but also for fission gas release and rod internal pressure, corrosion, dimensional stability. My colleagues and I have performed thousands of simulations for steady-state and transient (RIA and LOCA) analyses. Modeling a single fuel rod is complicated, let alone thousands of fuel rods (pins) in a core, e.g., 50952 fuel rods (x ~360 pellets per fuel rod) in a PWR core of 193 assemblies of 17x17 geometry, or a BWR with ~70288 to 76400 fuel rods in a core of 764 assemblies with 10x10 lattice geometry (we now have to consider 11x11 lattices). Modern BWR fuel designs have part length rods, in some case of two different lengths.
One has to consider the spatial and energy resolution (number of neutron energy groups), isotopic (radionuclide) resolution, thermohydraulic and temporal resolutions. Consider what one wishes to simulate, e.g., from a steady-state core depletion calculation, with normal startup and shutdowns and reduced power operation, to slow and fast transients. Some plants do considerable load following.
I've used results of 3D core simulations since years ago, but they were fairly coarse from today's more elaborate methods. The axial nodal resolution was 6 inches (152.4 mm).
The purpose of pin-by-pin core simulations without spatial homogenization is to accurately model the behavior of individual fuel pins in a nuclear reactor core. This approach takes into account the heterogeneity of materials within each pin, rather than homogenizing them into a single material. This allows for more detailed analysis and can provide more accurate results.
In traditional spatial homogenization methods, the materials within each fuel pin are homogenized into a single material with averaged properties. This simplifies the simulation but can lead to less accurate results. In pin-by-pin simulations, the heterogeneity of materials within each pin is taken into account, resulting in a more detailed and accurate simulation.
This approach can benefit any type of nuclear reactor that uses fuel pins, including light water reactors, fast breeder reactors, and advanced reactor designs. It is especially useful for reactors with complex fuel designs or highly heterogeneous materials.
One of the main challenges is the increased computational cost compared to traditional homogenization methods. This approach also requires detailed information about the materials and geometry of each fuel pin, which may not always be available. Additionally, the accuracy of the results depends on the quality of the input data and the assumptions made in the simulation.
By taking into account the heterogeneity of materials within each fuel pin, pin-by-pin simulations can provide more accurate predictions of reactor behavior, such as temperature distribution and fuel performance. This can lead to improved designs, increased safety, and better understanding of reactor performance. It also allows for more detailed sensitivity and uncertainty analyses, which can aid in decision-making for reactor operation and design.