The radiation characteristics of the transport packages with vitrified high-level waste

The calculation method of neutron yield in the (α, n ) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron...

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Published inPhysics of atomic nuclei Vol. 78; no. 11; pp. 1301 - 1308
Main Authors Bogatov, S. A., Mitenkova, E. F., Novikov, N. V.
Format Journal Article
LanguageEnglish
Published Moscow Pleiades Publishing 15.12.2015
Springer
Subjects
ALPHA REACTIONS
Analysis
CALCULATION METHODS
CASKS
HIGH-LEVEL RADIOACTIVE WASTES
NEUTRON FLUX
NEUTRONS
Nuclear energy
Nuclear industry
Nuclear radiation
NUCLEAR REACTION YIELD
O CODES
Particle and Nuclear Physics
Physics
Physics and Astronomy
Radioactive pollution
Reactor fuel reprocessing
SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
SPENT FUELS
Spent reactor fuels
VITRIFICATION
WWER TYPE REACTORS
radiation safety
high-level radioactive waste
transport cask
spent nuclear fuel
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Abstract The calculation method of neutron yield in the (α, n ) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n ) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.
AbstractList The calculation method of neutron yield in the (α, n ) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n ) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.
The calculation method of neutron yield in the (a, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (a, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified high-level waste are estimated. Keywords: spent nuclear fuel, high-level radioactive waste, transport cask, radiation safety. DOI: 10.1134/S106377881511006X
The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.
The calculation method of neutron yield in the (a, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (a, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified high-level waste are estimated.
Audience Academic
Author Mitenkova, E. F.
Bogatov, S. A.
Novikov, N. V.
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10.2172/5285077
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10.1007/BF01122564
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Issue 11
Keywords radiation safety
high-level radioactive waste
transport cask
spent nuclear fuel
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Snippet The calculation method of neutron yield in the (α, n ) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of...
The calculation method of neutron yield in the (a, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of...
The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of...
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SubjectTerms ALPHA REACTIONS
Analysis
CALCULATION METHODS
CASKS
HIGH-LEVEL RADIOACTIVE WASTES
NEUTRON FLUX
NEUTRONS
Nuclear energy
Nuclear industry
Nuclear radiation
NUCLEAR REACTION YIELD
O CODES
Particle and Nuclear Physics
Physics
Physics and Astronomy
Radioactive pollution
Reactor fuel reprocessing
SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
SPENT FUELS
Spent reactor fuels
VITRIFICATION
WWER TYPE REACTORS
Title The radiation characteristics of the transport packages with vitrified high-level waste
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https://www.osti.gov/biblio/22471952
Volume 78
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