Abstract
The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a `Heat Transfer package` is used for calculating heat transfer coefficient, DNB heat flux etc. The `Heat Transfer package` is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the `Heat Transfer package` and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author).
Kaminaga, Masanori
[1]
- Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
Citation Formats
Kaminaga, Masanori.
COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors.
Japan: N. p.,
1994.
Web.
Kaminaga, Masanori.
COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors.
Japan.
Kaminaga, Masanori.
1994.
"COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors."
Japan.
@misc{etde_10156057,
title = {COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors}
author = {Kaminaga, Masanori}
abstractNote = {The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a `Heat Transfer package` is used for calculating heat transfer coefficient, DNB heat flux etc. The `Heat Transfer package` is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the `Heat Transfer package` and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author).}
place = {Japan}
year = {1994}
month = {Mar}
}
title = {COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors}
author = {Kaminaga, Masanori}
abstractNote = {The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a `Heat Transfer package` is used for calculating heat transfer coefficient, DNB heat flux etc. The `Heat Transfer package` is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the `Heat Transfer package` and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author).}
place = {Japan}
year = {1994}
month = {Mar}
}