Zobrazeno 1 - 10
of 13
pro vyhledávání: '"Martin Lovecký"'
Publikováno v:
Acta Polytechnica CTU Proceedings, Vol 14, Iss 0, Pp 14-20 (2018)
In the depletion calculation of the nuclear fuel, the uncertainty is of utmost importance, as it affects the uncertainty of the subsequent calculation, when the calculated composition is used. The calculations are even more important when they are sa
Externí odkaz:
https://doaj.org/article/7ad0acf039d74e02bbfa0e31045da17a
Publikováno v:
Acta Polytechnica CTU Proceedings, Vol 4, Iss 0, Pp 43-49 (2016)
UWB1 depletion code is being developed as a fast computational tool for the study of burnable absorbers in University of West Bohemia in Pilsen, Czech Republic. Research of fuel depletion aims at development and introduction of advanced types of burn
Externí odkaz:
https://doaj.org/article/75c088acee9b46a1ad1e03123bd807e1
Publikováno v:
Kerntechnik. 85:336-340
This work deals with the updated reference solution of the Full-Core VVER-440 PK-3+ benchmark which is based on the extended calculation benchmark from 2018. [1] The modification consists of a new type of fuel assembly with PK-3 design in the loading
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::f3b60d1b9e55ad1412666d9f2cfdafbb
https://doi.org/10.3139/124.200020
https://doi.org/10.3139/124.200020
Autor:
Pavlína Haroková, Martin Lovecký
Publikováno v:
EPJ Web of Conferences, Vol 253, p 07011 (2021)
One of the methodologies used in criticality safety analysis is burnup credit method, which allows considering fuel burnup in models with spent fuel. This removes excessive conservatism from the analysis, but it also brings new uncertainties originat
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::0edfcf77bcac601295a2ad4a3987519e
https://www.epj-conferences.org/articles/epjconf/pdf/2021/07/epjconf_animma2021_07011.pdf
https://www.epj-conferences.org/articles/epjconf/pdf/2021/07/epjconf_animma2021_07011.pdf
The paper describes the development, validation, and use of UWB1 code intended for fast calculation of nuclear fuel depletion in burnable absorber research. The degree of effectiveness of burnable absorbers in the form of natural abundance elements,
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::e1f2acfaf7076528c1de79c0f140c5bf
http://hdl.handle.net/11025/36548
http://hdl.handle.net/11025/36548
Publikováno v:
EPJ Web of Conferences, Vol 247, p 17003 (2021)
Higher enrichment of nuclear fuel along the manufacturing limit of boron content in steel and aluminum alloys represents a significant challenge in designing spent fuel transport and storage facilities. One possible solution for spent fuel pools and
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::eb5b4105f4fcc80dd798e53d8711bba4
https://doi.org/10.1051/epjconf/202124717003
https://doi.org/10.1051/epjconf/202124717003
Publikováno v:
Progress in Nuclear Energy. 90:127-139
A computer code (UWB1) that is applicable for both light and heavy water reactor fuel types is being developed at the University of West Bohemia. CANDU fuel geometry was investigated in this paper. Fuel depletion with UWB1 code was performed and comp
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::fef401e9eaee6f23b79ed246736897a0
https://doi.org/10.1016/j.pnucene.2016.03.010
https://doi.org/10.1016/j.pnucene.2016.03.010
Autor:
P. Haroková, Jana Jiřičková, Martin Lovecký, Kristýna Klímek Gincelová, Jiří Závorka, Vladimír Smutný
Publikováno v:
Annals of Nuclear Energy. 148:107736
The latest ENDF/B nuclear data library released in 2018 is the result of a new international approach to develop evaluated nuclear data for general purpose applications. In order to use the latest ENDF/B-VIII.0 nuclear data library in safety analyses
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::affe94a3d9af59714af294a97ca1f1a5
https://doi.org/10.1016/j.anucene.2020.107736
https://doi.org/10.1016/j.anucene.2020.107736
Geometry models for Monte Carlo transport codes have been using standard constructive solid geometry (CSG). The standard approach is using analytical equations for defining surfaces from which spatial cells are constructed. Both union and intersectio
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::d905cb48a113701d90f13631a963ccaa
http://hdl.handle.net/11025/33833
http://hdl.handle.net/11025/33833
Publikováno v:
Annals of Nuclear Energy. 71:333-339
The research presented in this paper aims at the development and introduction of acomputational tool to evaluate advanced types of burnable absorbers (BA) in nuclear fuel. BAs compensate for the initial excess reactivity and consequently allow for lo
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::6549c689bf5f2036cabc685ebfdfd163
https://doi.org/10.1016/j.anucene.2014.04.021
https://doi.org/10.1016/j.anucene.2014.04.021