Evolution of irradiated Pile Grade A graphite microstructure under novel molten salt decontamination conditions

• Electrochemical decontamination has minimal effect on graphite microstructure • Oxygen pore diffusion and limited surface chlorination influence decontamination • Increasing current application results in creation of porosity • Stability of lattice parameters and crystallinity suggest no salt intercalation • Potential for waste volume reduction, re-categorisation, recycling or re-use The irradiated graphite waste stream represents a significant challenge for nuclear power plant decommissioning in the UK, with an estimated 96,000 tonnes of graphite waste arising from the shutdown of the UK's gas-cooled reactors. The removal of activated impurities and fission products from irradiated graphite has been successfully performed previously [ 1 ] using an electrochemical molten salt decontamination approach. In this study, the material behaviour and structural changes of the treated nuclear graphite under molten salt treatment conditions have been assessed using multi-technique characterisation (Brunauer–Emmett–Teller surface area, Scanning Electron Microscopy, X-ray Photoelectron Spectroscopy and X-ray Diffraction). This novel research highlights that molten salt treatment leads to changes to the binder and impregnated phases while leaving the filler particles intact under the researched treatment conditions. Significant differences in atomic concentrations of C 1s deconvoluted peaks were observed, suggesting that the mechanism involved both diffusion of pre-adsorbed oxygen and limited chlorination of the surface during the decontamination process. The stability of the lattice parameters and minimal change in crystalline dimensions in molten salt-treated graphite material combined with limited mass loss provides a first-of-a-kind insight into the mechanisms behind graphite decontamination using the electrochemical molten salt approach. The findings support the future potential for wide-scale irradiated graphite treatment to achieve waste volume reduction, minimisation and re-categorisation in line with the waste hierarchy.

Authors

Institutions

Publication Details

Journal
Journal of Nuclear Materials
Published
2025-05-30
DOI
https://doi.org/10.1016/j.jnucmat.2025.155935
Citations
3
Primary Topic
Graphite, nuclear technology, radiation studies
Type
article
Field-Weighted Citation Impact
1.55
Controls
|||
ALL TIME
JAN
FEB
MAR
APR
MAY
JUN
JUL
AUG
SEP
OCT
article

Evolution of irradiated Pile Grade A graphite microstructure under novel molten salt decontamination conditions

I. Ipatova, Daniel N.T. Barton, Robert N. Worth, Tatiana Grebennikova et al.
3 citations
Journal of Nuclear Materials
Graphite, nuclear technology, radiation studies
1.55
article

Evolution of irradiated Pile Grade A graphite microstructure under novel molten salt decontamination conditions

I. Ipatova, Daniel N.T. Barton, Robert N. Worth, Tatiana Grebennikova, Abbie Jones, Ben F. Spencer, Clint A. Sharrad
article en
3 citations

Abstract

• Electrochemical decontamination has minimal effect on graphite microstructure • Oxygen pore diffusion and limited surface chlorination influence decontamination • Increasing current application results in creation of porosity • Stability of lattice parameters and crystallinity suggest no salt intercalation • Potential for waste volume reduction, re-categorisation, recycling or re-use The irradiated graphite waste stream represents a significant challenge for nuclear power plant decommissioning in the UK, with an estimated 96,000 tonnes of graphite waste arising from the shutdown of the UK's gas-cooled reactors. The removal of activated impurities and fission products from irradiated graphite has been successfully performed previously [ 1 ] using an electrochemical molten salt decontamination approach. In this study, the material behaviour and structural changes of the treated nuclear graphite under molten salt treatment conditions have been assessed using multi-technique characterisation (Brunauer–Emmett–Teller surface area, Scanning Electron Microscopy, X-ray Photoelectron Spectroscopy and X-ray Diffraction). This novel research highlights that molten salt treatment leads to changes to the binder and impregnated phases while leaving the filler particles intact under the researched treatment conditions. Significant differences in atomic concentrations of C 1s deconvoluted peaks were observed, suggesting that the mechanism involved both diffusion of pre-adsorbed oxygen and limited chlorination of the surface during the decontamination process. The stability of the lattice parameters and minimal change in crystalline dimensions in molten salt-treated graphite material combined with limited mass loss provides a first-of-a-kind insight into the mechanisms behind graphite decontamination using the electrochemical molten salt approach. The findings support the future potential for wide-scale irradiated graphite treatment to achieve waste volume reduction, minimisation and re-categorisation in line with the waste hierarchy.

Journal of Nuclear MaterialsVol. 615
University of Manchester (GB), United Kingdom Atomic Energy Authority (GB)
Openalex Percentile: Top 18%
Graphite, nuclear technology, radiation studies
1.55
AI Navigator

Ask Laika to Summarize, Analyze, and Connect papers live on the map.

Summarize Papers & Methodologies

Extract key findings, datasets, and comparative methods across publications.

Benchmark Rankings & Visual Analytics

Rank top research institutions, authors, funders, topics, and journals by Field-Weighted Citation Impact (FWCI) and paper volume with instant charts.

Connect Distant Disciplines

Bridge topological clusters on the map to find hidden collaborative intersections.