Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code

Antti Rintala; Ville Valtavirta; Jaakko Leppänen

doi:10.1016/j.anucene.2021.108603

Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code

Antti Rintala^*
, Ville Valtavirta
, Jaakko Leppänen

^*Corresponding author for this work

Research output: Contribution to journal › Article › Scientific › peer-review

8 Citations (Scopus)

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Abstract

This paper describes the methodology used by the Monte Carlo particle transport code Serpent to produce homogenized microscopic cross sections. The microscopic cross sections can be used for multiple purposes for example in nodal programs. The current implementation is presented together with its features and limitations. The methodology is demonstrated by calculating the activation of structural materials in two simple test problems with Serpent and the nodal neutronics program Ants utilizing Serpent-generated group constants.

Original language	English
Article number	108603
Journal	Annals of Nuclear Energy
Volume	164
DOIs	https://doi.org/10.1016/j.anucene.2021.108603
Publication status	Published - 15 Dec 2021
MoE publication type	A1 Journal article-refereed

Funding

This work has received funding from the LONKERO project under The Finnish Research Programme on Nuclear Power Plant Safety 2019-2022 (SAFIR2022).

Keywords

Homogenization
Microscopic cross sections
Microscopic depletion
Serpent

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10.1016/j.anucene.2021.108603Licence: CC BY-NC-ND

j.anucene.2021.108603Final published version, 1.11 MBLicence: CC BY-NC-ND

Cite this

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title = "Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code",

abstract = "This paper describes the methodology used by the Monte Carlo particle transport code Serpent to produce homogenized microscopic cross sections. The microscopic cross sections can be used for multiple purposes for example in nodal programs. The current implementation is presented together with its features and limitations. The methodology is demonstrated by calculating the activation of structural materials in two simple test problems with Serpent and the nodal neutronics program Ants utilizing Serpent-generated group constants.",

keywords = "Homogenization, Microscopic cross sections, Microscopic depletion, Serpent",

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AU - Leppänen, Jaakko

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Y1 - 2021/12/15

N2 - This paper describes the methodology used by the Monte Carlo particle transport code Serpent to produce homogenized microscopic cross sections. The microscopic cross sections can be used for multiple purposes for example in nodal programs. The current implementation is presented together with its features and limitations. The methodology is demonstrated by calculating the activation of structural materials in two simple test problems with Serpent and the nodal neutronics program Ants utilizing Serpent-generated group constants.

AB - This paper describes the methodology used by the Monte Carlo particle transport code Serpent to produce homogenized microscopic cross sections. The microscopic cross sections can be used for multiple purposes for example in nodal programs. The current implementation is presented together with its features and limitations. The methodology is demonstrated by calculating the activation of structural materials in two simple test problems with Serpent and the nodal neutronics program Ants utilizing Serpent-generated group constants.

KW - Homogenization

KW - Microscopic cross sections

KW - Microscopic depletion

KW - Serpent

UR - https://www.scopus.com/pages/publications/85112278010

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DO - 10.1016/j.anucene.2021.108603

M3 - Article

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VL - 164

JO - Annals of Nuclear Energy

JF - Annals of Nuclear Energy

M1 - 108603

ER -

Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code

Abstract

Funding

Keywords

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