JOSS submission [do not merge]

.github/workflows/paper.yml

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+name: Draft PDF
+on: [push]
+
+jobs:
+  paper:
+    runs-on: ubuntu-latest
+    name: Paper Draft
+    steps:
+      - name: Checkout
+        uses: actions/checkout@v4
+      - name: Build draft PDF
+        uses: openjournals/openjournals-draft-action@master
+        with:
+          journal: joss
+          # This should be the path to the paper within your repo.
+          paper-path: paper.md
+      - name: Upload
+        uses: actions/upload-artifact@v4
+        with:
+          name: paper
+          # This is the output path where Pandoc will write the compiled
+          # PDF. Note, this should be the same directory as the input
+          # paper.md
+          path: paper.pdf

paper.bib

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+@article{fausser_tokamak_2012,
+	series = {Tenth {International} {Symposium} on {Fusion} {Nuclear} {Technology} ({ISFNT}-10)},
+	title = {Tokamak {D}-{T} neutron source models for different plasma physics confinement modes},
+	volume = {87},
+	issn = {0920-3796},
+	url = {https://www.sciencedirect.com/science/article/pii/S0920379612000853},
+	doi = {10.1016/j.fusengdes.2012.02.025},
+	abstract = {Neutronic studies of European demonstration fusion power plant (DEMO) have been so far based on plasma physics low confinement mode (L-mode). Future tokamaks, nevertheless, may likely use alternative confinement modes such as high or advanced confinement modes (H\&A-mode). Based on analytical formulae used in plasma physics, H\&A-modes D-T neutron sources formulae are proposed in this paper. For that purpose, a tokamak random neutron source generator, TRANSGEN, has been built generating bidimensional (radial and poloidal) neutron source maps to be used as input for neutronics Monte-Carlo codes (TRIPOLI-4 and MCNP5). The impact of such a source on the neutronic behavior of the European DEMO-2007 Helium-cooled lithium–lead reactor concept has been assessed and compared with previous results obtained using a L-mode neutron source. An A-mode neutron source map from TRANSGEN has been used with the code TRIPOLI-4. Assuming the same fusion power, results show that main reactor global neutronic parameters, e.g. tritium breeding ratio and neutron multiplication factor, evolved slightly when compared to present uncertainties margin. However, local parameters, such as the neutron wall loading (NWL), change significantly compared to L-mode shape: from −22\% to +11\% for NWL.},
+	language = {en},
+	number = {5},
+	urldate = {2021-09-21},
+	journal = {Fusion Engineering and Design},
+	author = {Fausser, Clement and Puma, Antonella Li and Gabriel, Franck and Villari, Rosaria},
+	month = aug,
+	year = {2012},
+	keywords = {DEMO, Neutronics, TRIPOLI-4, Plasma source description},
+	pages = {787--792},
+}
+
+@article{romano_openmc_2015,
+	series = {Joint {International} {Conference} on {Supercomputing} in {Nuclear} {Applications} and {Monte} {Carlo} 2013, {SNA} + {MC} 2013. {Pluri}- and {Trans}-disciplinarity, {Towards} {New} {Modeling} and {Numerical} {Simulation} {Paradigms}},
+	title = {{OpenMC}: {A} state-of-the-art {Monte} {Carlo} code for research and development},
+	volume = {82},
+	issn = {0306-4549},
+	shorttitle = {{OpenMC}},
+	url = {https://www.sciencedirect.com/science/article/pii/S030645491400379X},
+	doi = {10.1016/j.anucene.2014.07.048},
+	abstract = {This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes.},
+	language = {en},
+	urldate = {2022-03-31},
+	journal = {Annals of Nuclear Energy},
+	author = {Romano, Paul K. and Horelik, Nicholas E. and Herman, Bryan R. and Nelson, Adam G. and Forget, Benoit and Smith, Kord},
+	month = aug,
+	year = {2015},
+	keywords = {Monte Carlo, Parallel, HDF5, Neutron transport, OpenMC, XML},
+	pages = {90--97},
+	file = {ScienceDirect Snapshot:C\:\\Users\\remidm\\Zotero\\storage\\7Z6DVQXL\\S030645491400379X.html:text/html},
+}

paper.md

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+---
+title: 'openmc-plasma-source: Pre-built Fusion Neutron Sources for OpenMC'
+tags:
+  - Python
+  - OpenMC
+  - neutron sources
+  - fusion
+authors:
+  - name: Remi Delaporte-Mathurin
+    corresponding: true
+    orcid: 0000-0003-1064-8882
+    affiliation: 1
+  - name: Jonathan Shimwell
+    orcid: 0000-0001-6909-0946
+    affiliation: 2
+
+affiliations:
+ - name: Plasma Science and Fusion Center, MIT, USA
+   index: 1
+ - name: Proxima Fusion, Germany
+   index: 2
+
+date: 26 November 2024
+bibliography: paper.bib
+
+---
+
+# Summary
+
+`openmc-plasma-source` is a Python-based package offering a collection of pre-built neutron sources designed for fusion applications using the Monte Carlo particle transport code OpenMC [@romano_openmc_2015].
+By providing ready-to-use implementations for various neutron source configurations, such as tokamak, ring, and point sources, this package simplifies the often complex task of neutron source definition in fusion-related Monte Carlo simulations.
+These sources are parameterised to account for spatial distributions, plasma temperatures, and fusion fuel compositions, enabling realistic and computationally efficient neutron emission models.
+
+The package is designed to integrate seamlessly into OpenMC workflows, allowing users to define sources in just a few lines of Python code.
+It also supports advanced features like temperature-based neutron spectra and spatial source distributions, making it an invaluable tool for researchers simulating neutron behaviour in fusion devices.
+
+# Statement of need
+
+Accurate modelling of neutron sources is critical for fusion energy research, underpinning tasks such as reactor shielding design, material testing, and tritium breeding analysis.
+In this context, OpenMC is a widely used tool for neutron transport simulations [@romano_openmc_2015].
+However, creating realistic neutron source models for fusion applications can be a time-consuming and error-prone process, requiring detailed knowledge of plasma physics and significant coding effort.
+
+Traditionally, researchers have implemented their own custom neutron source definitions, which often results in redundant work and inconsistencies between studies. For example, spatial distributions, temperature effects, and fuel compositions must be correctly parameterised to ensure the fidelity of the simulations. The lack of standardised tools for these tasks introduces variability and potential errors in simulations.
+
+`openmc-plasma-source` addresses these challenges by providing a standardised and easy-to-use interface for defining neutron sources in OpenMC. The package implements the equations for neutron distributions based on established models, such as those described in @fausser_tokamak_2012. By automating the setup process and including extensive documentation and examples, it reduces barriers to entry for researchers new to OpenMC or neutron source modelling.
+
+With pre-built configurations for tokamak, ring, and point sources, `openmc-plasma-source` is suitable for a wide range of applications. For example:
+
+- The **tokamak source** models realistic spatial and temperature distributions, optimised for computational efficiency through the use of ring sources.
+- The **ring source** offers a simplified yet effective representation for cylindrical geometries.
+- The **point source** is ideal for preliminary studies or cases requiring a concentrated neutron emission found in inertial confinement fusion.
+
+
+The package’s open-source nature and community-driven development further ensure its adaptability and relevance to the evolving needs of the fusion research community.
+
+![Spatial neutron source density distributions of a tokamak source.\label{fig:tokamak_source}](https://user-images.githubusercontent.com/40028739/135100022-330aa51c-e2a2-401c-9738-90f3e99c84d4.png)
+
+![3D representation of a tokamak source.\label{fig:3d_tokamak_source}](3d_tokamak_source.png)
+
+# Example usage
+
+Examples can be found in the examples folder of the repository.
+
+# Acknowledgements
+
+We acknowledge contributions from the OpenMC development team and the fusion energy community for their feedback and support.
+
+# References