Abstract
A customized Computational Fluid Dynamics (CFD) model of supercritical water (SCW) heat transfer in a vertical tube upward flow is developed and partially validated using experimental data obtained under the operating conditions typical for SCW cooled reactors (SCWRs): at a pressure of 24 MPa, an inner tube diameter of 10 mm, an inlet temperature of 320 or 350 °C and a heated tube length of 4 m. The three values of mass flux (500, 1000 and 1500 kg/m2s) and various values of wall heat flux (from 141 to 729 kW/m2) are considered. Physical properties of SCW are calculated by using the REFPROP software from National Institute of Standards and Technology (NIST). The model has been incorporated into the commercial general-purpose CFD software, PHOENICS. Various turbulence models and numerical grid settings are tested. The study has demonstrated a good agreement between the CFD predictions and the experimental data on the inside tube wall temperature and heat transfer coefficient with use of a two-layer low-Reynolds-number k-ε turbulence model. However, a further model development is required under the conditions of significant effects of buoyancy force on heat transfer characteristics (the conditions of low values of mass flux and high values of wall heat flux). Practical recommendations are made regarding potential model applications in 3D analyses of SCWRs.
| Original language | English |
|---|---|
| Title of host publication | ICONE 2015 - 23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy |
| Publisher | American Society of Mechanical Engineers(ASME) |
| Volume | 2015-January |
| Publication status | Published - 2015 |
| Event | 23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015 - Chiba, Japan Duration: 17 May 2015 → 21 May 2015 |
Conference
| Conference | 23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015 |
|---|---|
| Country | Japan |
| City | Chiba |
| Period | 17.5.15 → 21.5.15 |
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Keywords
- Computational Fluid Dynamics
- Heat transfer
- PHOENICS software
- Supercritical Water
- Upward flow
- Validation
- Vertical tube
ASJC Scopus subject areas
- Nuclear Energy and Engineering
Cite this
CFD modeling of supercritical water heat transfer in a vertical bare tube upward flow. / Agranat, Vladimir; Malin, Michael; Pioro, Igor; Abdullah, Rand; Perminov, Valeriy Afanasievich.
ICONE 2015 - 23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy. Vol. 2015-January American Society of Mechanical Engineers(ASME), 2015.Research output: Chapter in Book/Report/Conference proceeding › Conference contribution
}
TY - GEN
T1 - CFD modeling of supercritical water heat transfer in a vertical bare tube upward flow
AU - Agranat, Vladimir
AU - Malin, Michael
AU - Pioro, Igor
AU - Abdullah, Rand
AU - Perminov, Valeriy Afanasievich
PY - 2015
Y1 - 2015
N2 - A customized Computational Fluid Dynamics (CFD) model of supercritical water (SCW) heat transfer in a vertical tube upward flow is developed and partially validated using experimental data obtained under the operating conditions typical for SCW cooled reactors (SCWRs): at a pressure of 24 MPa, an inner tube diameter of 10 mm, an inlet temperature of 320 or 350 °C and a heated tube length of 4 m. The three values of mass flux (500, 1000 and 1500 kg/m2s) and various values of wall heat flux (from 141 to 729 kW/m2) are considered. Physical properties of SCW are calculated by using the REFPROP software from National Institute of Standards and Technology (NIST). The model has been incorporated into the commercial general-purpose CFD software, PHOENICS. Various turbulence models and numerical grid settings are tested. The study has demonstrated a good agreement between the CFD predictions and the experimental data on the inside tube wall temperature and heat transfer coefficient with use of a two-layer low-Reynolds-number k-ε turbulence model. However, a further model development is required under the conditions of significant effects of buoyancy force on heat transfer characteristics (the conditions of low values of mass flux and high values of wall heat flux). Practical recommendations are made regarding potential model applications in 3D analyses of SCWRs.
AB - A customized Computational Fluid Dynamics (CFD) model of supercritical water (SCW) heat transfer in a vertical tube upward flow is developed and partially validated using experimental data obtained under the operating conditions typical for SCW cooled reactors (SCWRs): at a pressure of 24 MPa, an inner tube diameter of 10 mm, an inlet temperature of 320 or 350 °C and a heated tube length of 4 m. The three values of mass flux (500, 1000 and 1500 kg/m2s) and various values of wall heat flux (from 141 to 729 kW/m2) are considered. Physical properties of SCW are calculated by using the REFPROP software from National Institute of Standards and Technology (NIST). The model has been incorporated into the commercial general-purpose CFD software, PHOENICS. Various turbulence models and numerical grid settings are tested. The study has demonstrated a good agreement between the CFD predictions and the experimental data on the inside tube wall temperature and heat transfer coefficient with use of a two-layer low-Reynolds-number k-ε turbulence model. However, a further model development is required under the conditions of significant effects of buoyancy force on heat transfer characteristics (the conditions of low values of mass flux and high values of wall heat flux). Practical recommendations are made regarding potential model applications in 3D analyses of SCWRs.
KW - Computational Fluid Dynamics
KW - Heat transfer
KW - PHOENICS software
KW - Supercritical Water
KW - Upward flow
KW - Validation
KW - Vertical tube
UR - http://www.scopus.com/inward/record.url?scp=84959052444&partnerID=8YFLogxK
UR - http://www.scopus.com/inward/citedby.url?scp=84959052444&partnerID=8YFLogxK
M3 - Conference contribution
AN - SCOPUS:84959052444
VL - 2015-January
BT - ICONE 2015 - 23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy
PB - American Society of Mechanical Engineers(ASME)
ER -