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FORAL, Š. KATOVSKÝ, K. SUK, L. POLÁŠEK, R.
Original Title
VALIDATION OF THE ALTHAMC12 SUBCHANNEL CODE
Type
conference paper
Language
English
Original Abstract
One of the tools for safety analysis of light water reactors is the subchannel methodology which is used for evaluation of the limits with regards to departure from nucleate boiling (DNB). This approach utilities subchannel code which has to be validated against experimental data. In this work, the ALTHAMC12 subchannel code is validated against critical heat flux data measured at test section with hexagonal geometry. Validation is performed using the PG-S and OKB correlations of critical heat flux (CHF). The experimental data library is based on public literature sources. The analysis shows that PG-S correlation provides prediction of CHF close to experimental values. The OKB correlation underestimates the CHF and thus provides conservative results for safety analysis.
Keywords
Subchanel code; nuclear thermal hydraulics; DNBR
Authors
FORAL, Š.; KATOVSKÝ, K.; SUK, L.; POLÁŠEK, R.
Released
11. 11. 2021
Publisher
American Nuclear Society
Location
555 North Kensington Avenue La Grange Park, Illinois 60526 U.S.A.
ISBN
9780894487774
Book
ATH 2020 - International Topical Meeting on Advances in Thermal Hydraulics
Edition
Proceedings of ANS
Edition number
1
Pages from
768
Pages to
781
Pages count
13
URL
https://www.ans.org/pubs/proceedings/article-49145/
BibTex
@inproceedings{BUT175919, author="Štěpán {Foral} and Karel {Katovský} and Ladislav {Suk} and Radek {Polášek}", title="VALIDATION OF THE ALTHAMC12 SUBCHANNEL CODE", booktitle="ATH 2020 - International Topical Meeting on Advances in Thermal Hydraulics", year="2021", series="Proceedings of ANS", number="1", pages="768--781", publisher="American Nuclear Society", address="555 North Kensington Avenue La Grange Park, Illinois 60526 U.S.A.", isbn="9780894487774", url="https://www.ans.org/pubs/proceedings/article-49145/" }