Foreseen capabilities, bottlenecks identification and potential limitations of serpent MC transport code in large-scale full 3-D burnup calculations

Diego Ferraro, Manuel Garcia, Luigi Mercatali, Victor Hugo Sanchez Espinoza, Jaakko Leppänen, Ville Valtavirta

Research output: Chapter in Book/Report/Conference proceedingConference article in proceedingsScientificpeer-review

2 Citations (Scopus)

Abstract

Continuous improvement in Nuclear Industry Safety Standards and reactor designer’s and operator’s commercial goals lead to an increasing demand of fast running and highly accurate methodologies oriented to improve the prediction capabilities of main reactor’s parameters under steady state and transient situations. An increasing effort has been observed during past years to develop high accurate multi-physics approach for nuclear reactor analysis, based both on the availability of advanced codes and the constant increase of computational resources with massive parallel architectures. As a result, several improvements have been observed in the implementation of coupled full 3-D Monte Carlo (MC) neutronic models for nuclear reactor cores, including not only the coupling to thermal-hydraulics but also fuel behavior codes. This approach has proved at concept level to be able to develop high accurate models that would allow to predict important safety and performance parameters of nuclear reactors with less conservativism. Under this framework, the European Research and innovation project McSAFE is a coordinated effort started in September 2017 with the objective of moving MC stand-alone and coupled solution methodologies to become valuable tools for core design, safety analysis and industry like applications for LWRs of gen II and III. In this work the Serpent 2 code, a high performance MC code developed by VTT, is used by the aim of performing the preliminary screening of capabilities, performance and limitations of such challenging objective. As a result, simplified analysis are developed to identify full 3-D modeling computational requirements for typical LWR configurations, including burnup aspects. Potential bottlenecks and limitations are presented and discussed, providing foreseen alternatives and solutions for further code improvements.

Original languageEnglish
Title of host publication26th International Conference on Nuclear Engineering
Subtitle of host publicationNuclear Fuel and Material, Reactor Physics, and Transport Theory
PublisherAmerican Society of Mechanical Engineers ASME
Number of pages9
Volume3
ISBN (Print)978-0-7918-5145-6
DOIs
Publication statusPublished - Nov 2018
MoE publication typeNot Eligible
Event26th International Conference on Nuclear Engineering, ICONE 2018 - London, United Kingdom
Duration: 22 Jul 201826 Jul 2018
Conference number: 26

Conference

Conference26th International Conference on Nuclear Engineering, ICONE 2018
Abbreviated titleICONE 2018
CountryUnited Kingdom
CityLondon
Period22/07/1826/07/18

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Nuclear reactors
Nuclear industry
Parallel architectures
Reactor cores
Screening
Physics
Innovation
Hydraulics
Availability
Industry
Hot Temperature

Cite this

Ferraro, D., Garcia, M., Mercatali, L., Sanchez Espinoza, V. H., Leppänen, J., & Valtavirta, V. (2018). Foreseen capabilities, bottlenecks identification and potential limitations of serpent MC transport code in large-scale full 3-D burnup calculations. In 26th International Conference on Nuclear Engineering: Nuclear Fuel and Material, Reactor Physics, and Transport Theory (Vol. 3). [ICONE26-82305] American Society of Mechanical Engineers ASME. https://doi.org/10.1115/ICONE26-82305
Ferraro, Diego ; Garcia, Manuel ; Mercatali, Luigi ; Sanchez Espinoza, Victor Hugo ; Leppänen, Jaakko ; Valtavirta, Ville. / Foreseen capabilities, bottlenecks identification and potential limitations of serpent MC transport code in large-scale full 3-D burnup calculations. 26th International Conference on Nuclear Engineering: Nuclear Fuel and Material, Reactor Physics, and Transport Theory. Vol. 3 American Society of Mechanical Engineers ASME, 2018.
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abstract = "Continuous improvement in Nuclear Industry Safety Standards and reactor designer’s and operator’s commercial goals lead to an increasing demand of fast running and highly accurate methodologies oriented to improve the prediction capabilities of main reactor’s parameters under steady state and transient situations. An increasing effort has been observed during past years to develop high accurate multi-physics approach for nuclear reactor analysis, based both on the availability of advanced codes and the constant increase of computational resources with massive parallel architectures. As a result, several improvements have been observed in the implementation of coupled full 3-D Monte Carlo (MC) neutronic models for nuclear reactor cores, including not only the coupling to thermal-hydraulics but also fuel behavior codes. This approach has proved at concept level to be able to develop high accurate models that would allow to predict important safety and performance parameters of nuclear reactors with less conservativism. Under this framework, the European Research and innovation project McSAFE is a coordinated effort started in September 2017 with the objective of moving MC stand-alone and coupled solution methodologies to become valuable tools for core design, safety analysis and industry like applications for LWRs of gen II and III. In this work the Serpent 2 code, a high performance MC code developed by VTT, is used by the aim of performing the preliminary screening of capabilities, performance and limitations of such challenging objective. As a result, simplified analysis are developed to identify full 3-D modeling computational requirements for typical LWR configurations, including burnup aspects. Potential bottlenecks and limitations are presented and discussed, providing foreseen alternatives and solutions for further code improvements.",
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Ferraro, D, Garcia, M, Mercatali, L, Sanchez Espinoza, VH, Leppänen, J & Valtavirta, V 2018, Foreseen capabilities, bottlenecks identification and potential limitations of serpent MC transport code in large-scale full 3-D burnup calculations. in 26th International Conference on Nuclear Engineering: Nuclear Fuel and Material, Reactor Physics, and Transport Theory. vol. 3, ICONE26-82305, American Society of Mechanical Engineers ASME, 26th International Conference on Nuclear Engineering, ICONE 2018, London, United Kingdom, 22/07/18. https://doi.org/10.1115/ICONE26-82305

Foreseen capabilities, bottlenecks identification and potential limitations of serpent MC transport code in large-scale full 3-D burnup calculations. / Ferraro, Diego; Garcia, Manuel; Mercatali, Luigi; Sanchez Espinoza, Victor Hugo; Leppänen, Jaakko; Valtavirta, Ville.

26th International Conference on Nuclear Engineering: Nuclear Fuel and Material, Reactor Physics, and Transport Theory. Vol. 3 American Society of Mechanical Engineers ASME, 2018. ICONE26-82305.

Research output: Chapter in Book/Report/Conference proceedingConference article in proceedingsScientificpeer-review

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Ferraro D, Garcia M, Mercatali L, Sanchez Espinoza VH, Leppänen J, Valtavirta V. Foreseen capabilities, bottlenecks identification and potential limitations of serpent MC transport code in large-scale full 3-D burnup calculations. In 26th International Conference on Nuclear Engineering: Nuclear Fuel and Material, Reactor Physics, and Transport Theory. Vol. 3. American Society of Mechanical Engineers ASME. 2018. ICONE26-82305 https://doi.org/10.1115/ICONE26-82305