Friction and Contact Mechanics of Nuclear Graphite in High-Temperature Gas-Cooled Reactors: A Review

Abstract

The high-temperature gas-cooled reactor (HTGR), as a preferred reactor type for Generation IV nuclear energy systems, widely employs nuclear-grade graphite in its core as the moderator, structural material, and matrix of fuel elements. The friction coefficient between graphite components and the contact stiffness of spherical fuel elements directly affect the structural integrity of the core, the flow characteristics of the pebble bed, and the design of the fuel handling system, serving as critical mechanical parameters for ensuring reactor safety and economic operation. This paper systematically reviews the preparation processes and multi-scale microstructural features of nuclear graphite for HTGRs and summarizes the current research status in two directions: the friction coefficient and the contact stiffness of nuclear graphite. In the field of tribology, the influencing mechanisms of intrinsic factors such as grain size and porosity, as well as extrinsic factors such as ambient atmosphere, temperature, load, and sliding velocity, on the friction behavior of graphite are analyzed with emphasis. The controversies and applicable conditions of the surface energy mechanism and the dangling bond mechanism are discussed. In terms of contact mechanics, the development of asperity models from the Hertz contact and Abbott-Firestone (AF) model to the Kogut-Etsion finite-element based fitting is summarized, along with the evolution from the Greenwood-Williamson (GW) statistical model to the Majumdar-Bhushan (MB) fractal contact model. Experimental methods for measuring stiffness coefficients- namely, the interface displacement method, the impact method, and the acoustic method-are reviewed. On this basis, the severe scarcity of nuclear graphite mechanical data under in-core conditions, such as high temperatures and inert atmospheres, is highlighted. Two systematic experimental works dedicated to graphite materials for HTGRs are integrated: the first measurements of the static and dynamic friction coefficients of isostatically pressed nuclear graphite BG80 in a high-purity helium environment from 25 to 1300 °C, revealing a four-stage variation of the friction coefficient with temperature; and the first acquisition of the temperature dependence and irreversible evolution characteristics of the stiffness coefficient of spherical fuel elements in high temperature helium. These works fill the gaps in fundamental data and can provide quantitative parameter inputs and theoretical support for HTGR core design, pebble-bed flow simulations, and safety analyses.

Keywords

High temperature gas cooled reactor / Nuclear graphite / Friction coefficient / Contact stiffness / Fuel element / Fractal contact model

Author Contributions

Conceptualization, T.L.; Methodology, M.S. and Z.C.; Writing-Original Draft Preparation, T.L.; Supervision, N.G.; Project Administration, S.J.

Ethics Statement

Not applicable.

Informed Consent Statement

Not applicable.

Data Availability Statement

Data will be available from the corresponding author on request.

Funding

This research was funded by [National Major Science and Technology Program, and LingChuang Research Project of China National Nuclear Corporation].

Declaration of Competing Interest

The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.

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