Prospects for the Use of TiO₂ to Improve the Properties of Cement Matrix during Encapsulation of Irradiated Graphite from Nuclear Facilities

O. M. Skoblyk

Kyiv Academic University, 36, Akademika Vernadskoho Blvd., Kyiv, 03142, Ukraine

https://orcid.org/0009-0006-0822-6428

K. V. Simeiko

Institute for Safety Problems of Nuclear Power Plants of the NAS of Ukraine, 36a, Kirova St., Chornobyl, 07270, Ukraine

The Gas Institute of the NAS of Ukraine, 39, Dehtiarivska St., Kyiv, 03113, Ukraine

https://orcid.org/0000-0002-1590-6281

DOI: doi.org/10.31717/2311-8253.26.2.5

Abstract

Since the early development of the nuclear industry, graphite has been widely used as a neutron moderator and reflector in various types of reactors (RBMK, Magnox, AGR, HTGR), resulting in the accumulation of over 250,000 t of irradiated graphite (IG) worldwide. In Ukraine, approximately 5,700 t of IG from RBMK‑1000 reactors at the Chornobyl NPP require management. No unified industrial solution for IG management currently exists. Cement immobilisation remains the most technologically accessible and internationally recognised conditioning method. This paper analyses the prospects of modifying cement matrices with TiO₂ nanoparticles for IG immobilisation. The fundamental limitation of the two-component Cement–IG system is the hydrophobic graphite surface, which creates a defective interfacial transition zone (ITZ). The ITZ increases matrix diffusion permeability and reduces barrier properties against mobile radionuclides ¹⁴C, ³H, and ³⁶Cl. An additional degradation factor is the radiation-mechanical destruction of C–S–H phases and radiation-chemical radiolysis of pore solution under γand β-irradiation. The introduction of TiO₂ nanoparticles addresses these limitations through three mechanisms: formation of Ti–O–C chemical bonds on the carbon surface (ITZ compensation); microstructure densification through the nano-filler effect (reduction in water absorption by 37.5−81.3 %, improvement in abrasion resistance by 8.2−58 %); enhancement of γ-shielding properties through increased linear attenuation coefficient μ and potential limitation of β-induced radiolysis of pore solution through the semiconducting properties of TiO₂. A critical research gap is identified: no published study has examined the ternary Cement–TiO₂–IG system with evaluation of radionuclide leaching under irradiation and long-term stability according to radioactive waste acceptance criteria. Priority research directions are defined: verification of the Ti–O–C bonding mechanism at the matrix–graphite interface, optimisation of the Cement–TiO₂–IG system composition for maximum microstructural density and minimum radionuclide leaching, and assessment of the radiation-chemical stability of the matrix according to radioactive waste acceptance protocols.

Keywords: irradiated graphite, encapsulation, immobilisation, cementitious matrix, titanium dioxide, γ-radiation, β-radiation, graphite-moderated nuclear reactors.

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