The interaction of molecular hydrogen with a-radiolytic oxidants on a (U,Pu)O2 surface

Research output: Contribution to journalArticle

Standard

The interaction of molecular hydrogen with a-radiolytic oxidants on a (U,Pu)O2 surface. / Bauhn, Lovisa; Hansson, Niklas; Ekberg, Christian; Fors, Patrik; Delville, Rémi; Spahiu, Kastriot.

In: Journal of Nuclear Materials, Vol. 505, 05.04.2018, p. 54-61.

Research output: Contribution to journalArticle

Harvard

Bauhn, L, Hansson, N, Ekberg, C, Fors, P, Delville, R & Spahiu, K 2018, 'The interaction of molecular hydrogen with a-radiolytic oxidants on a (U,Pu)O2 surface', Journal of Nuclear Materials, vol. 505, pp. 54-61. https://doi.org/10.1016/j.jnucmat.2018.04.006

APA

Bauhn, L., Hansson, N., Ekberg, C., Fors, P., Delville, R., & Spahiu, K. (2018). The interaction of molecular hydrogen with a-radiolytic oxidants on a (U,Pu)O2 surface. Journal of Nuclear Materials, 505, 54-61. https://doi.org/10.1016/j.jnucmat.2018.04.006

Vancouver

Bauhn L, Hansson N, Ekberg C, Fors P, Delville R, Spahiu K. The interaction of molecular hydrogen with a-radiolytic oxidants on a (U,Pu)O2 surface. Journal of Nuclear Materials. 2018 Apr 5;505:54-61. https://doi.org/10.1016/j.jnucmat.2018.04.006

Author

Bauhn, Lovisa ; Hansson, Niklas ; Ekberg, Christian ; Fors, Patrik ; Delville, Rémi ; Spahiu, Kastriot. / The interaction of molecular hydrogen with a-radiolytic oxidants on a (U,Pu)O2 surface. In: Journal of Nuclear Materials. 2018 ; Vol. 505. pp. 54-61.

Bibtex - Download

@article{761e6b1b0a8842558db98fe5e3b37e75,
title = "The interaction of molecular hydrogen with a-radiolytic oxidants on a (U,Pu)O2 surface",
abstract = "In order to assess the impact of a-radiolysis of water on the oxidative dissolution of spent fuel, an unirradiated, annealed MOX fuel pellet with high content of Pu (~24 wt{\%}), and a specific a-activity of 4.96 GBq/gMOX, was leached in carbonate-containing solutions of low ionic strength. The high Pu content in the pellet stabilizes the (U,Pu)O2(s) matrix towards oxidative dissolution, whereas the a-decays emitted from the surface are expected to produce ~3.6 107 mol H2O2/day, contributing to the oxidative dissolution of the pellet. Two sets of leaching tests were conducted under different redox conditions: Ar gas atmosphere and deuterium gas atmosphere. A relatively slow increase of the U and Pu concentrations was observed in the Ar case, with U concentrations increasing from 1$10-6M after 1 h to ~7 10-5M after 58 days. Leaching under an atmosphere starting at 1MPa deuterium gas was undertaken in order to evaluate any effect of dissolved hydrogen on the radiolytic dissolution of the pellet, as well as to investigate any potential recombination of the a-radiolytic products with dissolved deuterium. For the latter purpose, isotopic analysis of the D/H content was carried out on solution samples taken during the leaching. Despite the continuous production of radiolytic oxidants, the concentrations of U and Pu remained quite constant at the level of ~3 10-8M during the first 30 days, i.e. as long as the deuterium pressure remained higher than 0.8 MPa. These data rule out any oxidative dissolution of the pellet during the first month. The un-irradiated MOX fuel does not contain metallic ε-particles, hence it is mainly the interaction of radiolytic oxidants and dissolved deuterium with the surface of the mixed actinide oxide that causes the neutralization of the oxidants. This conclusion is supported by the steadily increasing levels of HDO measured in the leachate samples.",
keywords = "MOX, leaching, radiolysis, hydrogen, oxidants, uranium, plutonium",
author = "Lovisa Bauhn and Niklas Hansson and Christian Ekberg and Patrik Fors and R{\'e}mi Delville and Kastriot Spahiu",
note = "Score=10",
year = "2018",
month = "4",
day = "5",
doi = "10.1016/j.jnucmat.2018.04.006",
language = "English",
volume = "505",
pages = "54--61",
journal = "Journal of Nuclear Materials",
issn = "0022-3115",
publisher = "Elsevier",

}

RIS - Download

TY - JOUR

T1 - The interaction of molecular hydrogen with a-radiolytic oxidants on a (U,Pu)O2 surface

AU - Bauhn, Lovisa

AU - Hansson, Niklas

AU - Ekberg, Christian

AU - Fors, Patrik

AU - Delville, Rémi

AU - Spahiu, Kastriot

N1 - Score=10

PY - 2018/4/5

Y1 - 2018/4/5

N2 - In order to assess the impact of a-radiolysis of water on the oxidative dissolution of spent fuel, an unirradiated, annealed MOX fuel pellet with high content of Pu (~24 wt%), and a specific a-activity of 4.96 GBq/gMOX, was leached in carbonate-containing solutions of low ionic strength. The high Pu content in the pellet stabilizes the (U,Pu)O2(s) matrix towards oxidative dissolution, whereas the a-decays emitted from the surface are expected to produce ~3.6 107 mol H2O2/day, contributing to the oxidative dissolution of the pellet. Two sets of leaching tests were conducted under different redox conditions: Ar gas atmosphere and deuterium gas atmosphere. A relatively slow increase of the U and Pu concentrations was observed in the Ar case, with U concentrations increasing from 1$10-6M after 1 h to ~7 10-5M after 58 days. Leaching under an atmosphere starting at 1MPa deuterium gas was undertaken in order to evaluate any effect of dissolved hydrogen on the radiolytic dissolution of the pellet, as well as to investigate any potential recombination of the a-radiolytic products with dissolved deuterium. For the latter purpose, isotopic analysis of the D/H content was carried out on solution samples taken during the leaching. Despite the continuous production of radiolytic oxidants, the concentrations of U and Pu remained quite constant at the level of ~3 10-8M during the first 30 days, i.e. as long as the deuterium pressure remained higher than 0.8 MPa. These data rule out any oxidative dissolution of the pellet during the first month. The un-irradiated MOX fuel does not contain metallic ε-particles, hence it is mainly the interaction of radiolytic oxidants and dissolved deuterium with the surface of the mixed actinide oxide that causes the neutralization of the oxidants. This conclusion is supported by the steadily increasing levels of HDO measured in the leachate samples.

AB - In order to assess the impact of a-radiolysis of water on the oxidative dissolution of spent fuel, an unirradiated, annealed MOX fuel pellet with high content of Pu (~24 wt%), and a specific a-activity of 4.96 GBq/gMOX, was leached in carbonate-containing solutions of low ionic strength. The high Pu content in the pellet stabilizes the (U,Pu)O2(s) matrix towards oxidative dissolution, whereas the a-decays emitted from the surface are expected to produce ~3.6 107 mol H2O2/day, contributing to the oxidative dissolution of the pellet. Two sets of leaching tests were conducted under different redox conditions: Ar gas atmosphere and deuterium gas atmosphere. A relatively slow increase of the U and Pu concentrations was observed in the Ar case, with U concentrations increasing from 1$10-6M after 1 h to ~7 10-5M after 58 days. Leaching under an atmosphere starting at 1MPa deuterium gas was undertaken in order to evaluate any effect of dissolved hydrogen on the radiolytic dissolution of the pellet, as well as to investigate any potential recombination of the a-radiolytic products with dissolved deuterium. For the latter purpose, isotopic analysis of the D/H content was carried out on solution samples taken during the leaching. Despite the continuous production of radiolytic oxidants, the concentrations of U and Pu remained quite constant at the level of ~3 10-8M during the first 30 days, i.e. as long as the deuterium pressure remained higher than 0.8 MPa. These data rule out any oxidative dissolution of the pellet during the first month. The un-irradiated MOX fuel does not contain metallic ε-particles, hence it is mainly the interaction of radiolytic oxidants and dissolved deuterium with the surface of the mixed actinide oxide that causes the neutralization of the oxidants. This conclusion is supported by the steadily increasing levels of HDO measured in the leachate samples.

KW - MOX

KW - leaching

KW - radiolysis

KW - hydrogen

KW - oxidants

KW - uranium

KW - plutonium

UR - http://ecm.sckcen.be/OTCS/llisapi.dll/open/31287381

U2 - 10.1016/j.jnucmat.2018.04.006

DO - 10.1016/j.jnucmat.2018.04.006

M3 - Article

VL - 505

SP - 54

EP - 61

JO - Journal of Nuclear Materials

JF - Journal of Nuclear Materials

SN - 0022-3115

ER -

ID: 4614975