Evaluation of neutron activation of intermetallic matrices for dispersive nuclear fuel obtained by SH-synthesis

Abstract

The paper describes a technique for obtaining intermetallic matrix materials based on Zr-Al and Ni-Al systems for dispersive nuclear fuel. This type of fuel is planned to be used in existing and future high-temperature nuclear reactors. Physical and mathematical modelling of the process of activation of matrices by neutron radiation showed that upon reaching the value of thermal neutron fluence of the order of 2.76⋅10²¹ n⋅cm^-2, the activity of Zr-Al and Ni-Al matrices was 1.3 ⋅ 10¹⁰ and 0.2⋅10¹⁰ Bq/g, respectively. The analysis showed that in terms of exposure of irradiated fuel it is preferable to use a matrix based on the zirconium-aluminium system.

Original language	English
Article number	012010
Journal	Journal of Physics: Conference Series
Volume	1439
Issue number	1
DOIs	https://doi.org/10.1088/1742-6596/1439/1/012010
Publication status	Published - 17 Jan 2020
Event	8th International Youth Scientific School-Conference on Modern Problems of Physics and Technology, MPPT 2019 - Moscow, Russian Federation Duration: 15 Apr 2019 → 20 Apr 2019

ASJC Scopus subject areas

Physics and Astronomy(all)

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10.1088/1742-6596/1439/1/012010

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Evaluation of neutron activation of intermetallic matrices for dispersive nuclear fuel obtained by SH-synthesis. / Balachov, M. M.; Dolmatov, O. Yu ; Kuznetsov, M. S.; Kostyuchenko, K. S.; Pimenov, N. O.; Permikin, A. A.

In: Journal of Physics: Conference Series, Vol. 1439, No. 1, 012010, 17.01.2020.

Research output: Contribution to journal › Conference article › peer-review

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title = "Evaluation of neutron activation of intermetallic matrices for dispersive nuclear fuel obtained by SH-synthesis",

abstract = "The paper describes a technique for obtaining intermetallic matrix materials based on Zr-Al and Ni-Al systems for dispersive nuclear fuel. This type of fuel is planned to be used in existing and future high-temperature nuclear reactors. Physical and mathematical modelling of the process of activation of matrices by neutron radiation showed that upon reaching the value of thermal neutron fluence of the order of 2.76⋅1021 n⋅cm-2, the activity of Zr-Al and Ni-Al matrices was 1.3 ⋅ 1010 and 0.2⋅1010 Bq/g, respectively. The analysis showed that in terms of exposure of irradiated fuel it is preferable to use a matrix based on the zirconium-aluminium system.",

author = "Balachov, {M. M.} and Dolmatov, {O. Yu} and Kuznetsov, {M. S.} and Kostyuchenko, {K. S.} and Pimenov, {N. O.} and Permikin, {A. A.}",

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T1 - Evaluation of neutron activation of intermetallic matrices for dispersive nuclear fuel obtained by SH-synthesis

AU - Balachov, M. M.

AU - Dolmatov, O. Yu

AU - Kuznetsov, M. S.

AU - Kostyuchenko, K. S.

AU - Pimenov, N. O.

AU - Permikin, A. A.

PY - 2020/1/17

Y1 - 2020/1/17

N2 - The paper describes a technique for obtaining intermetallic matrix materials based on Zr-Al and Ni-Al systems for dispersive nuclear fuel. This type of fuel is planned to be used in existing and future high-temperature nuclear reactors. Physical and mathematical modelling of the process of activation of matrices by neutron radiation showed that upon reaching the value of thermal neutron fluence of the order of 2.76⋅1021 n⋅cm-2, the activity of Zr-Al and Ni-Al matrices was 1.3 ⋅ 1010 and 0.2⋅1010 Bq/g, respectively. The analysis showed that in terms of exposure of irradiated fuel it is preferable to use a matrix based on the zirconium-aluminium system.

AB - The paper describes a technique for obtaining intermetallic matrix materials based on Zr-Al and Ni-Al systems for dispersive nuclear fuel. This type of fuel is planned to be used in existing and future high-temperature nuclear reactors. Physical and mathematical modelling of the process of activation of matrices by neutron radiation showed that upon reaching the value of thermal neutron fluence of the order of 2.76⋅1021 n⋅cm-2, the activity of Zr-Al and Ni-Al matrices was 1.3 ⋅ 1010 and 0.2⋅1010 Bq/g, respectively. The analysis showed that in terms of exposure of irradiated fuel it is preferable to use a matrix based on the zirconium-aluminium system.

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