Speaker
Description
Nuclear fusion is pursued with distinct methodologies, including magnetic confinement fusion, inertial confinement fusion, and inertial electrostatic confinement fusion (IECF) [1]. The neutron and alpha particle emissions characteristic of D-T fusion exhibit dependency on fuel selection, with compositional modifications directly influencing resultant reaction outputs [2, 3]. As documented in the IEC book, the neutron flux generation rate of IEC device is $10^{6}$ to $10^{12}$ $\mathrm{s}^{-1}$ [4]. Consequently, radiation shielding optimization for this neutron-emitting apparatus must incorporate spectral considerations of fusion-generated neutrons and secondary particle emissions to ensure operational safety. Prior studies have established frameworks for shielding design and radiological hazard mitigation in such fusion systems [5, 6]. Among the investigated shielding materials for IECF devices, prior research has employed structures comprising paraffin, boric acid (H3BO3), wood, and stainless steel to achieve effective radiation attenuation [5]. Our previous research [7,8] successfully implemented shield design simulations for this system, demonstrating effective radiation mitigation capabilities.
But in this work, by simulating with the MCNPX code, the attenuation performance of B$_4$C and W layers was analyzed for D-T fusion neutron shielding. Boron carbide (B$_4$C) is a well-established thermal neutron absorber due to its high neutron capture cross-section [9], while tungsten (W) exhibits superior gamma-ray attenuation properties owing to its high atomic density [10]. This work evaluates the substitution of B$_4$C for boric acid (H$_3$BO$_3$) and W for lead (Pb) in multilayer shielding configurations to optimize neutron and gamma radiation attenuation for a 14.1 MeV deuterium-tritium (D-T) fusion neutron source. We simulated an IECF device with a neutron source strength of $10^{9}$ $\mathrm{s}^{-1}$ and isotropic angular emission with the MCNPX code. The distribution of energy is assumed as a Gaussian energy, defined by the 14.1 MeV neutron spectrum. Computational rigor was ensured by tracking $2\times 10^{6}$ particle histories (nps), resulting in statistical uncertainties below 1%. Neutron and gamma fluxes were quantified using F2 surface tallies, with dose calculations employing the DFn card (IU=2 configuration), standardizing dose units to $\mathrm{Sv}\cdot \mathrm{h}^{-1}$source$^{-1}$. Comparative analysis of shielding materials revealed that B4C-W configuration achieved gamma dose reduction (1.23 μSv), whereas H3BO3-W configuration demonstrated enhanced neutron attenuation (15.40 μSv). Gamma dosimetry results (1.23 μSv) were minimized using the W layer, while using H$_3$BO$_3$ with the W layer proved optimal for reducing neutron dose and flux (15.40 μSv). Material composition and material effects composition at fixed geometries reveal critical trade-offs in optimization and multifunctional radiation materials design. These results show a good improvement in reducing neutron and gamma doses compared to the results of previous studies, whose articles have also been published.
References
1. J. Black et al., Physical Review E 103(2), 023212 (2021).
2. Y.A. Chan et al., Vacuum 167, 482-489 (2019).
3. P.T. Farnsworth, No. US 3258402 (1966).
4. G.H Miley & S.K. Murali (2014). Inertial electrostatic confinement (IEC) fusion. Fundam. Appl.
5. S.M Lee et al., International Nuclear Atlantic Conference, Santos, Brazil 5088–5095 (2020).
6. M. Bakr et al., Radiation Research and Applied Sciences 17(2), 100908 (2024).
7. H. Zanganeh & M. Nasri Nasrabadi, Radiation Physics and Engineering 4(3), 29-41 (2023).
8. H. Zanganeh, & M.N. Nasrabadi, Radiation Physics and Chemistry 229, p.112495 (2025).
9. E. Mansouri et al., International Journal of Radiation Research 18(4), 611-622 (2020).
10. M. Asgari et al., Nuclear Engineering and Technology 53(12), 4142-4149 (2021).