Molten Salt Tritium Breeding Materials in Fusion Reactors: A Neutronic Comparative Analysis for ITER

New Energy Exploitation and Application

Article

Molten Salt Tritium Breeding Materials in Fusion Reactors: A Neutronic Comparative Analysis for ITER

Downloads

https://doi.org/10.54963/neea.v4i2.1413
Karakoç, A. (2025). Molten Salt Tritium Breeding Materials in Fusion Reactors: A Neutronic Comparative Analysis for ITER. New Energy Exploitation and Application, 4(2), 144–156. https://doi.org/10.54963/neea.v4i2.1413
Volume 4 Issue 2 (2025): In Progress
Copyright (c) 2025 Alper Karakoç
Creative Commons License

This work is licensed under a Creative Commons Attribution 4.0 International License.

Copyright

The authors shall retain the copyright of their work but allow the Publisher to publish, copy, distribute, and convey the work.

License

New Energy Exploitation and Application (NEEA)  publishes accepted manuscripts under Creative Commons Attribution 4.0 International (CC BY 4.0). Authors who submit their papers for publication by New Energy Exploitation and Application (NEEA) agree to have the CC BY 4.0 license applied to their work, and that anyone is allowed to reuse the article or part of it free of charge for any purpose, including commercial use. As long as the author and original source is properly cited, anyone may copy, redistribute, reuse and transform the content.

Authors

  • Alper Karakoç

    Independent Researcher, Ankara 06824, Türkiye

Received: 15 July 2025; Revised: 27 August 2025; Accepted: 2 September 2025; Published: 25 September 2025

This study investigates the effects of varying tritium breeding materials and their lithium enrichment rates on the Tritium Breeding Ratio (TBR) and Energy Multiplication Factor (M) within the tritium breeding zone of a fusion reactor. The magnetic fusion reactor model was developed based on the geometric and plasma parameters of the International Thermonuclear Experimental Reactor (ITER), ensuring a realistic representation of current fusion reactor designs. ITER‑grade stainless steel (SS 316 LN‑IG) was selected as the first wall material due to its excellent mechanical properties, high resistance to radiation damage, and compatibility with high‑temperature environments. The coolant and tritium breeding materials considered in the blanket included natural lithium, lithium fluoride (LiF), FLiBe (LiF‑BeF2), and FLiNaBe (LiF‑NaF‑BeF2). These materials were chosen for their ability to facilitate tritium breeding while maintaining thermal and neutronic efficiency. Neutron transport calculations and geometric modeling were performed using the widely recognized 3D simulation tools MCNP 5 and TopMC, which employ the continuous‑energy Monte Carlo method. The simulations utilized built‑in continuous‑energy nuclear and atomic data libraries, along with the Evaluated Nuclear Data File (ENDF) system (ENDF/B‑V and ENDF/B‑VI), ensuring reliable and validated results. The results highlight the importance of material selection and enrichment optimization in achieving efficient tritium breeding and energy production. FLiBe, in particular, shows promise for future fusion reactor designs due to its superior performance in terms of TBR and M. These findings provide valuable insights for the development of sustainable and high‑performance fusion reactors, contributing to the global pursuit of clean and virtually limitless energy.

Keywords:

ITER Magnetic Fusion Reactor Tritium Breeding Ratio Energy Multiplication Factor Neutronic Analysis

References

  1. Eddington, A.S. The Internal Constitution of the Stars. Cambridge University Press: New York, NY, USA, 1988.
  2. Greenberger, D.; Hentschel, K.; Weinert, F. Compendium of Quantum Physics: Concepts, Experiments, History and Philosophy. Springer: Berlin, Germany, 2009.
  3. Sakharov, A. Memoirs. Alfred A. Knopf: New York, NY, USA, 1990.
  4. Post, R.F. Summary of UCRL Pyrotron Programme. J. Nucl. Energy 1958, 7, 282.
  5. El-Guebaly, L.A. Fifty Years of Magnetic Fusion Research (1958–2008): Brief Historical Overview and Discussion of Future Trends. Energies 2010, 3, 1067–1086.
  6. ITER Organization. Available online: https://www.iter.org (accessed on 2 June 2025).
  7. Neilson, G. Magnetic Fusion Energy: From Experiments to Power Plants, 1st ed. Woodhead Publishing: Cambridge, UK, 2016.
  8. Federici, G.; Biel, W.; Gilbert, M.R.; et al. European DEMO Design Strategy and Consequences for Materials. Nucl. Fusion 2017, 57, 9. DOI: https://doi.org/10.1088/1741-4326/57/9/092002
  9. Barabaschi, P.; Fossen, A.; Loarte, A.; et al. ITER Progresses into New Baseline. Fusion Eng. Des. 2025, 215, 114990. DOI: https://doi.org/10.1016/j.fusengdes.2025.114990
  10. Romanelli, F.; Paméla, J.; Kamendje, R.; et al. Recent Contribution of JET to the ITER Physics. Fusion Eng. Des. 2009, 84, 150–160. DOI: https://doi.org/10.1016/j.fusengdes.2008.11.088
  11. Li, J.; Wan, Y.; Wan, B.; et al. Lessons Learned from EAST’s Failures. In Proceedings of the 24th IAEA Fusion Energy Conference, San Diego, CA, USA, 8–13 October 2012.
  12. Abdou, M.; Morley, N.B.; Smolentsev, S.; et al. Blanket/First Wall Challenges and Required R&D on the Pathway to DEMO. Fusion Eng. Des. 2015, 100, 2–43. DOI: https://doi.org/10.1016/j.fusengdes.2015.07.021
  13. Zinkle, S.J.; Snead, L.L. Designing Radiation Resistance in Materials for Fusion Energy. Annu. Rev. Mater. Res. 2014, 44, 241–267. DOI: https://doi.org/10.1146/annurev-matsci-070813-113627
  14. Liu, S.; Li, J.; Zheng, S.; et al. Neutronics Analysis of Inboard Shielding Capability for a DEMO Fusion Reactor CFETR. Fusion Eng. Des. 2013, 88, 2404–2407. DOI: https://doi.org/10.1016/j.fusengdes.2013.04.031
  15. Boullon, R.; Jaboulay, J.C.; Aubert, J. Molten Salt Breeding Blanket: Investigations and Proposals of Pre-Conceptual Design Options for Testing in DEMO. Fusion Eng. Des. 2021, 171, 112707. DOI: https://doi.org/10.1016/j.fusengdes.2021.112707
  16. Arp, V.D.; McCarty, R.D. Thermophysical Properties of Helium-4 from 0.8 to 1500 K with Pressures to 2000 MPa. U.S. Government Printing Office: Washington, DC, USA, 1989.
  17. Cismondi, F.; Boccaccini, L.V.; Aiello, G.; et al. Progress in EU Breeding Blanket Design and Integration. Fusion Eng. Des. 2018, 136, 782–792. DOI: https://doi.org/10.1016/j.fusengdes.2018.04.009
  18. Arbeiter, F.; Chen, Y.; Ghidersa, B.E.; et al. Options for a High Heat Flux Enabled Helium Cooled First Wall for DEMO. Fusion Eng. Des. 2017, 119, 22–28. DOI: https://doi.org/10.1016/j.fusengdes.2017.04.083
  19. Iguchi, T.; Sekiguchi, A.; Nakazawa, M. A Benchmark Experiment on Tritium Production and Radiation Heating in the LIF Assembly. Nucl. Technol. Fusion 1983, 4, 817–822. DOI: https://doi.org/10.13182/fst83-a22961
  20. Şahin, S. Selection Criteria For Fusion Reactor Structures. J. Therm. Eng. 2019, 5, 46–57.
  21. Krat, S.A.; Popkov, A.S.; Gasparyan, Y.M.; et al. Wetting Properties of Liquid Lithium on Lithium Compounds. Fusion Eng. Des. 2017, 117, 199–203. DOI: https://doi.org/10.1016/j.fusengdes.2016.06.038
  22. Palermo, I.; Gómez-Ros, J.M.; Veredas, G.; et al. Neutronic Design Analyses for a Dual-Coolant Blanket Concept: Optimization for a Fusion Reactor DEMO. Fusion Eng. Des. 2012, 87, 1019–1024. DOI: https://doi.org/10.1016/j.fusengdes.2012.02.067
  23. Beaufait, R.; Fischer, L. Blanket Cooling of a Fusion Reactor. Energies 2023, 16, 1890.
  24. Forsberg, C.; Zheng, G.; Ballinger, R.G.; et al. Fusion Blankets and Fluoride-Salt-Cooled High-Temperature Reactors with Flibe Salt Coolant: Common Challenges, Tritium Control, and Opportunities for Synergistic Development Strategies Between Fission, Fusion, and Solar Salt Technologies. Nucl. Technol. 2020, 206, 1778–1801.
  25. Nygren, R.E.; Rognlien, T.D.; Rensink, M.E.; et al. A Fusion Reactor Design with a Liquid First Wall and Divertor. Fusion Eng. Des. 2004, 72, 181–221.
  26. Sawan, M.; Abdou, M. Physics and Technology Conditions for Attaining Tritium Self-Sufficiency for the DT Fuel Cycle. Fusion Eng. Des. 2006, 81, 1131–1144.
  27. Schreinlechner, I.; Šmid, I.; Weimann, G. Evaluation of SS 316 First Wall Material for Activation and Afterheat for Maintenance Purposes. Fusion Eng. Des. 1991, 17, 403–407. DOI: https://doi.org/10.1016/0920-3796(91)90087-7
  28. Zheng, S.; Todd, T.N. Study of Impacts on Tritium Breeding Ratio of a Fusion DEMO Reactor. Fusion Eng. Des. 2015, 98, 1915–1918.
  29. Fukada, S.; Edao, Y.; Sagara, A. Effects of Simultaneous Transfer of Heat and Tritium through Li–Pb or Flibe Blanket. Fusion Eng. Des. 2010, 85, 1314–1319. DOI: https://doi.org/10.1016/j.fusengdes.2010.03.032
  30. Fischer, U.; Pereslavtsev, P.; Grosse, D.; et al. Nuclear Design Analyses of the Helium Cooled Lithium Lead Blanket for a Fusion Power Demonstration Reactor. Fusion Eng. Des. 2010, 85, 1133–1138. DOI: https://doi.org/10.1016/j.fusengdes.2010.02.023
  31. Boccaccini, L.V.; Fiek, H.J.; Fischer, U.; et al. Advanced Helium Cooled Pebble Bed Blanket. Forschungszentrum Karlsruhe GmbH: Karlsruhe, Germany, 2000.
  32. Pandey, S.K.; Samal, M.K. Experimental Evaluation of Temperature and Strain-Rate-Dependent Mechanical Properties of Austenitic Stainless Steel SS 316 LN and a New Methodology to Evaluate Parameters of Johnson–Cook and Ramberg–Osgood Material Models. Solids 2025, 6, 7. DOI: https://doi.org/10.3390/solids6010007
  33. De Temmerman, G.; Heinola, K.; Borodin, D.; et al. Data on Erosion and Hydrogen Fuel Retention in Beryllium Plasma-Facing Materials. Nucl. Mater. Energy 2021, 27, 100994. DOI: https://doi.org/10.1016/j.nme.2021.100994
  34. Hirai, T.; Bao, L.; Barabash, V.; et al. High Heat Flux Performance Assessment of ITER Enhanced Heat Flux First Wall Technology after Neutron Irradiation. Fusion Eng. Des. 2023, 186, 113338. DOI: https://doi.org/10.1016/j.fusengdes.2022.113338
  35. Rieth, M.; Dudarev, S.L.; de Vicente, S.M.G.; et al. Recent Progress in Research on Tungsten Materials for Nuclear Fusion Applications in Europe. J. Nucl. Mater. 2013, 432, 482–500. DOI: https://doi.org/10.1016/j.jnucmat.2012.08.018
  36. Federici, G.; Skinner, C.H.; Brooks, J.N.; et al. Plasma–Material Interactions in Current Tokamaks and Their Implications for Next Step Fusion Reactors. Nucl. Fusion 2001, 41, 12. DOI: https://doi.org/10.1088/0029-5515/41/12/218
  37. Mitteau, R.; Calcagno, B.; Chappuis, P.; et al. The Design of the ITER First Wall Panels. Fusion Eng. Des. 2013, 88, 568–570. DOI: https://doi.org/10.1016/j.fusengdes.2013.05.030
  38. Hernández, F.A.; Pereslavtsev, P. First Principles Review of Options for Tritium Breeder and Neutron Multiplier Materials for Breeding Blankets in Fusion Reactors. Fusion Eng. Des. 2018, 137, 243–256. DOI: https://doi.org/10.1016/j.fusengdes.2018.09.014
  39. Santoro, R.T. Radiation Shielding for Fusion Reactors. J. Nucl. Sci. Technol. 2000, 37, 11–18.
  40. X-5 Monte Carlo Team. MCNP – A General Monte Carlo N‑Particle Transport Code, Version 5: Volume I. Overview and Theory (LA‑UR‑03‑1987). Los Alamos National Laboratory: Los Alamos, NM, USA, 2003.
  41. Chen, S.; Wu, B.; Dong, Z.; et al. Development of TopMC 1.0 for Nuclear Technology Applications. EPJ Nucl. Sci. Technol. 2025, 11, 6. DOI: https://doi.org/10.1051/epjn/2024033
  42. Maemunah, I.; Rosidah, I.; Su’ud, Z.; et al. Tritium Breeding Performance Analysis of HCLL Blanket Fusion Reactor Employing Vanadium Alloy (V-5Cr-5Ti) as First Wall Material. Sci. Technol. Nucl. Install. 2022, 2022, 5300160. DOI: https://doi.org/10.1155/2022/5300160
  43. Park, J.H.; Pereslavtsev, P.; Konobeev, A.; et al. Statistical Analysis of Tritium Breeding Ratio Deviations in the DEMO Due to Nuclear Data Uncertainties. Appl. Sci. 2021, 11, 5234. DOI: https://doi.org/10.3390/app11115234
  44. Chen, H.; Pan, L.; Lv, Z.; et al. Tritium Fuel Cycle Modeling and Tritium Breeding Analysis for CFETR. Fusion Eng. Des. 2016, 106, 17–20. DOI: https://doi.org/10.1016/j.fusengdes.2016.02.100
  45. Jolodosky, A.; Fratoni, M. Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets. U.S. Department of Energy (DOE) through Lawrence Livermore National Laboratory: Livermore, CA, USA, 2015.
  46. Şahin, S.; Şarer, B.; Çelik, Y. Energy Multiplication and Fissile Fuel Breeding Limits of Accelerator-Driven Systems with Uranium and Thorium Targets. Int. J. Hydrogen Energy 2015, 40, 4037–4046. DOI: https://doi.org/10.1016/j.ijhydene.2015.01.141
  47. Şahin, H.M.; Tunç, G.; Karakoç, A. Neutronic Analysis of a Tokamak Hybrid Blanket Cooled by Thorium-Molten Salt Fuel Mixture. Prog. Nucl. Energy 2024, 174, 105306. DOI: https://doi.org/10.1016/j.pnucene.2024.105306