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Neutronics in Fusion

Fusion Reactions

The principal nuclear fusion reactions of interest in fusion energy are summarized below, showing their reaction channels and total energy release (Q-value).

 
D–T Nuclear Fusion Reaction. Source: Neutron Sources, [1]


Reaction Q-value (MeV)
2H+3H⟶4He+n (14.1 MeV) 17.6
2H+2H⟶3He+n (2.45 MeV) 3.27
2H+2H⟶3H+p 4.03
2H+3He⟶4He+p 18.3
3He+3He⟶4He+2p 12.86
p+11B⟶34He 8.68


Neutron Interactions in Fusion Devices

Fusion neutrons escape the plasma and interact with reactor materials, producing heat and influencing material behavior. The principal neutron interactions in fusion reactors are summarized below:

Interaction Effect
Radiation Damage Causes atomic displacements and transmutation (He, H), degrading mechanical properties such as strength and ductility
Activation Transforms stable materials into radioactive isotopes, affecting maintenance, waste management, and radiation exposure
Tritium Breeding Neutrons react with lithium to produce tritium [1]:
  • 6Li+n⟶4He+3H(+4.78 MeV)
  • 7Li+n⟶4He+3H+n(−2.47 MeV)


Note: Because tritium is extremely rare in nature (half-life ≈ 12.3 years), fusion reactors are initially supplied with tritium from existing stockpiles (primarily from CANDU fission reactors), after which tritium is continuously bred from lithium within the reactor blanket to sustain operation[2].


Neutron Modeling in Plasma Codes

The following plasma modeling codes are used to predict fast-ion behavior and fusion-born neutron emission in tokamak and stellarator plasmas, providing neutron source distributions for transport simulations and diagnostic design.

Plasma Code Description and Typical Applications
TRANSP Models plasma conditions and fast-ion dynamics in tokamaks; outputs are used as neutron source terms for transport simulations [3][4]
FIDASIM Simulates fast-ion behavior and predicts fusion-born neutron emission; used to estimate signals for neutron diagnostics like cameras and spectrometers [5]
NUBEAM Module of TRANSP that models fast-ion distributions from neutral beam injection and fusion reactions; computes neutron source profiles [6]
ASCOT Monte Carlo orbit-following codes for fast ions in tokamaks and stellarators; predict localized neutron emission patterns to aid diagnostic design and calibration [7][8]


Neutron Modeling Using Monte Carlo Simulations

Monte Carlo neutron transport codes are widely used in fusion research to model neutron behavior, material interactions, and diagnostic responses. The table below summarizes the principal neutron-related work performed by each code, together with representative references.

Code Neutron-Related Work in Fusion
MCNP Neutron transport from plasma to reactor components; shielding design; nuclear heating; activation analysis; neutron diagnostics response modeling [9]
Serpent Neutron transport and spectral analysis in fusion blankets; tritium breeding ratio (TBR) calculations; activation and decay heat studies [10]
PHITS Coupled neutron and charged-particle transport; detailed 3D neutronics; radiation damage indicators (DPA, gas production); shielding studies [11]
OpenMC High-fidelity neutron transport in complex 3D fusion geometries; neutron diagnostics scoping; activation analysis; uncertainty and sensitivity studies [12]


Useful GitHub Repositories for Neutronics in Nuclear Fusion

Several GitHub repositories provide tools and workflows that are highly valuable for neutronics development in fusion research:

  1. fusion-energy – This is a massive and highly recommended GitHub organization for any neutron physicist working in fusion. It hosts a wide range of projects for fusion neutronics and contains tools like fusion_neutron_utils, fusion_neutronics_workflow, and OpenMC plasma source utilities. These projects enable neutron source generation, Monte Carlo transport simulations, activation and decay analysis, and 3D modeling of fusion device neutronics, making it an indispensable resource for both research and development (GitHub: https://github.com/fusion-energy).
  2. aurora-multiphysics / aurora – Integrates neutron transport with multiphysics solvers, allowing simulation of neutron energy deposition, radiation effects, and thermal feedback in complex reactor geometries. (GitHub: https://github.com/aurora-multiphysics/aurora)
  3. Fusion-Power-Plant-Framework / tokamak-neutron-source – Provides Python tools to generate parametric tokamak neutron sources for Monte Carlo transport codes such as OpenMC and MCNP, based on plasma profiles and fusion reaction distributions. (GitHub: https://github.com/Fusion-Power-Plant-Framework/tokamak-neutron-source)


References

  1. ↑ "Breeding blanket," *Wikipedia*, https://en.wikipedia.org/wiki/Breeding_blanket
  2. ↑ "Tritium production in CANDU reactors," *Wikipedia*, https://en.wikipedia.org/wiki/Tritium#Production
  3. ↑ A. Sperduti et al., “Validation of neutron emission and neutron energy spectrum calculations on a Mega Ampere Spherical Tokamak,” *Plasma Physics and Controlled Fusion*, vol. 63, no. 1, 2021.
  4. ↑ “Generation of a plasma neutron source for Monte Carlo neutron transport calculations in the tokamak JET,” *Fusion Engineering and Design*, vol. 136, 2018.
  5. ↑ B. Geiger, L. Stagner, W. W. Heidbrink et al., “Progress in modelling fast‑ion D‑alpha spectra and neutral particle analyzer fluxes using FIDASIM,” *Plasma Physics and Controlled Fusion*, 2020.
  6. ↑ A. Pankin et al., “The tokamak Monte Carlo fast ion module NUBEAM in the National Transport Code Collaboration library,” *Computer Physics Communications*, vol. 159, 2004.
  7. ↑ H. Weisen, P. Sirén, J. Varje & JET Contributors, “Comparison of JET‑C DD neutron rates independently predicted by the ASCOT and TRANSP Monte Carlo heating codes,” *Nuclear Fusion*, vol. 62, 016017, 2022.
  8. ↑ J. Kontula et al., “ASCOT simulations of 14 MeV neutron rates in W7‑X: effect of magnetic configuration,” arXiv:2009.02925, 2020.
  9. ↑ C. J. Werner (ed.), MCNP User’s Manual – Code Version 6.2, Los Alamos National Laboratory, LA-UR-17-29981.
  10. ↑ J. Leppänen et al., The Serpent Monte Carlo Code: Status, Development and Applications in Fusion Neutronics, Annals of Nuclear Energy.
  11. ↑ T. Sato et al., Features of Particle and Heavy Ion Transport Code System (PHITS), Journal of Nuclear Science and Technology.
  12. ↑ P. K. Romano et al., OpenMC: A State-of-the-Art Monte Carlo Code for Research and Development, Annals of Nuclear Energy.
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