Decontamination of Irradiated Nuclear Graphite Using High-Temperature Molten Salt-17447Citation formats

Standard

Decontamination of Irradiated Nuclear Graphite Using High-Temperature Molten Salt-17447. / Grebennikova, Tatiana; Sharrad, Clint; Jones, Abbie.

2017. 5560-5567 Paper presented at 43rd Annual Waste Management Conference, Phoenix, United States.

Research output: Contribution to conferencePaper

Harvard

Grebennikova, T, Sharrad, C & Jones, A 2017, 'Decontamination of Irradiated Nuclear Graphite Using High-Temperature Molten Salt-17447' Paper presented at 43rd Annual Waste Management Conference, Phoenix, United States, 5/03/17 - 9/03/17, pp. 5560-5567.

APA

Grebennikova, T., Sharrad, C., & Jones, A. (2017). Decontamination of Irradiated Nuclear Graphite Using High-Temperature Molten Salt-17447. 5560-5567. Paper presented at 43rd Annual Waste Management Conference, Phoenix, United States.

Vancouver

Grebennikova T, Sharrad C, Jones A. Decontamination of Irradiated Nuclear Graphite Using High-Temperature Molten Salt-17447. 2017. Paper presented at 43rd Annual Waste Management Conference, Phoenix, United States.

Author

Bibtex

@conference{be93bed713b746719a79b303c562bad9,
title = "Decontamination of Irradiated Nuclear Graphite Using High-Temperature Molten Salt-17447",
abstract = "Irradiated graphite is one of the most significant, large volume waste streams in the UK. After shut down of gas cooled reactors there will be ~ 96,000 tonnes of nuclear graphite arising from pressure vessels, sealed or unsealed stacks, temporary surface storage and in silos, which may account for up to 30{\%} by ILW volume of any future UK geological disposal facility. High temperature molten salt treatment (HMST) could be considered as a partitioning process of the activation and fission products from irradiated graphite. The main objective of the research is to optimize a specific graphite treatment technology compatible with older, current and future reactors to provide a safe and effective process to decontaminate graphite and reduce the waste inventory. In order to reach that purpose principal radionuclides contained in irradiated samples from Magnox reactors were investigated by germanium (Ge) gamma spectrometry. The next key tasks were to employ treatment at 450ºC in LiCl-KCl eutectic which included the following procedures: electrochemical cleaning of the salt, initial cyclic voltammetry (CV) of graphite followed by several steps of chronopotentiometry (CP) with different current applied and CV was taken after each step. Once the treatment was established, experiments investigating the electrorefinement of the resultant salt mixtures were conducted as well as the comparison of gamma spectroscopy results of graphite before and after the treatment providing the result of total activity reduction of around 60{\%}.",
author = "Tatiana Grebennikova and Clint Sharrad and Abbie Jones",
year = "2017",
language = "English",
pages = "5560--5567",
note = "43rd Annual Waste Management Conference : Education & Opportunity in Waste Management, WM2017 ; Conference date: 05-03-2017 Through 09-03-2017",

}

RIS

TY - CONF

T1 - Decontamination of Irradiated Nuclear Graphite Using High-Temperature Molten Salt-17447

AU - Grebennikova, Tatiana

AU - Sharrad, Clint

AU - Jones, Abbie

PY - 2017

Y1 - 2017

N2 - Irradiated graphite is one of the most significant, large volume waste streams in the UK. After shut down of gas cooled reactors there will be ~ 96,000 tonnes of nuclear graphite arising from pressure vessels, sealed or unsealed stacks, temporary surface storage and in silos, which may account for up to 30% by ILW volume of any future UK geological disposal facility. High temperature molten salt treatment (HMST) could be considered as a partitioning process of the activation and fission products from irradiated graphite. The main objective of the research is to optimize a specific graphite treatment technology compatible with older, current and future reactors to provide a safe and effective process to decontaminate graphite and reduce the waste inventory. In order to reach that purpose principal radionuclides contained in irradiated samples from Magnox reactors were investigated by germanium (Ge) gamma spectrometry. The next key tasks were to employ treatment at 450ºC in LiCl-KCl eutectic which included the following procedures: electrochemical cleaning of the salt, initial cyclic voltammetry (CV) of graphite followed by several steps of chronopotentiometry (CP) with different current applied and CV was taken after each step. Once the treatment was established, experiments investigating the electrorefinement of the resultant salt mixtures were conducted as well as the comparison of gamma spectroscopy results of graphite before and after the treatment providing the result of total activity reduction of around 60%.

AB - Irradiated graphite is one of the most significant, large volume waste streams in the UK. After shut down of gas cooled reactors there will be ~ 96,000 tonnes of nuclear graphite arising from pressure vessels, sealed or unsealed stacks, temporary surface storage and in silos, which may account for up to 30% by ILW volume of any future UK geological disposal facility. High temperature molten salt treatment (HMST) could be considered as a partitioning process of the activation and fission products from irradiated graphite. The main objective of the research is to optimize a specific graphite treatment technology compatible with older, current and future reactors to provide a safe and effective process to decontaminate graphite and reduce the waste inventory. In order to reach that purpose principal radionuclides contained in irradiated samples from Magnox reactors were investigated by germanium (Ge) gamma spectrometry. The next key tasks were to employ treatment at 450ºC in LiCl-KCl eutectic which included the following procedures: electrochemical cleaning of the salt, initial cyclic voltammetry (CV) of graphite followed by several steps of chronopotentiometry (CP) with different current applied and CV was taken after each step. Once the treatment was established, experiments investigating the electrorefinement of the resultant salt mixtures were conducted as well as the comparison of gamma spectroscopy results of graphite before and after the treatment providing the result of total activity reduction of around 60%.

M3 - Paper

SP - 5560

EP - 5567

ER -