Effects of axial power shapes on CHF locations in a single tube and in rod bundle assemblies
-
Abstract
Currently, the prediction of rod bundle CHF is dependent on CHF correlations derived from CHF data. A simple correction factor, such as F-factor, is often used to account for the axial power shape differences based on a simple accumulated energy concept, which has totally no consideration on the impact of true local condition on CHF mechanism. Subsequently, as expected, large uncertainty is often associated with the CHF value and CHF location predictions. For the purpose of obtaining different power shapes effects on CHF, CFD calculated parameter values were used to predict the possible CHF occurrence location. The possible CHF location prediction method proposed in this paper is calculated void fraction, heat transfer coefficient (HTC), liquid temperature distribution and detailed local parameters. And the uniform and non-uniform CHF were analyzed. The prediction of possible CHF locations in a 5 × 5 rod bundle may provide useful information for the design of a full-length CHF test, enhance the accuracy of CHF and CHF location prediction, and reduce the costs of the experimentation.
Kurzfassung
Die Berechnung der kritischen Heizflächenbelastung in einem Stabbündel basiert derzeit auf aus Experimenten abgeleiteten Korrelationen. Oft wird in diese ein sog. Korrekturfaktor (F-Faktor) eingeführt, um axiale Leistungsverteilungen zu berücksichtigen. Dieser berücksichtigt jedoch nie die tatsächlichen lokalen Bedingungen. So sind dann die daraus berechneten Werte für den kritischen Wärmestrom und den Ort seines Auftretens mit großen Unsicherheiten behaftet. Um den Einfluss unterschiedlicher axialer Leistungsverteilungen auf das Auftreten der kritischen Heizflächenbelastung zu untersuchen, wurden CFD Analysen durchgeführt. Die Ergebnisse werden in diesem Beitrag vorgestellt. Dabei werden basierend auf der Berechnung des Dampfgehalts, der Wärmeübergangskoeffizienten, der Flüssigkeitstemperaturverteilung und weiterer lokaler Größen. Es werden gleichmäßige und nicht gleichmäßige Leistungsverteilungen berücksichtigt. Die Ergebnisse erster Rechnungen mit einem 5 × 5 Stabbündel werden vorgestellt.
References
1 Yang, B.; Shan,J.; Gou,J.; et al.: Uniform versus Nonuniform Axial Power Distribution in Rod Bundle CHF Experiments. Science and Technology of Nuclear Installations2014, 201410.1155/2014/462460Search in Google Scholar
2 Zhang, R.; Cong, T.; Tian, W.; et al.: CFD analysis on subcooled boiling phenomena in PWR coolant channel. Progress in Nuclear Energy81 (2015) 254–26310.1016/j.pnucene.2015.02.005Search in Google Scholar
3 Tong, L. S.; Tang, Y. S.: Boiling heat transfer and two-phase flow. CRC press, 1997Search in Google Scholar
4 Liu, W.; Nariai, H.: Viewpoint of subcooled flow boiling critical heat flux mechanism. Chemical Engineering & Technology25 (2002) 447–45310.1002/1521-4125(200204)25:4<447::AID-CEAT447>3.0.CO;2-PSearch in Google Scholar
5 CU-HTRF-2001-W1010, Critical Heat Flux Tests on PWR Fuel Assemblies for Westinghouse Electric Company Test No. 101.0, 2001Search in Google Scholar
6 Glück, M.: Sub-channel analysis with F-COBRA-TF-Code validation and approaches to CHF prediction. Nuclear Engineering and Design237 (2007) 655–66710.1016/j.nucengdes.2006.08.011Search in Google Scholar
7 Mimouni, S.; Baudry, C.; Guingo, M.; et al.: Computational multi-fluid dynamics predictions of critical heat flux in boiling flow. Nuclear Engineering and Design299 (2016) 28–3610.1016/j.nucengdes.2015.07.017Search in Google Scholar
8 Kurul, N.; Podowski, M. Z.: On the modeling of multidimensional effects in boiling channels. ANS Proceeding of the 27th National Heat Transfer Conference. 1991Search in Google Scholar
9 Weisman, J.; Pei, B. S.: Prediction of critical heat flux in flow boiling at low qualities. International Journal of Heat and Mass Transfer26 (1983) 1463–147710.1016/S0017-9310(83)80047-7Search in Google Scholar
10 Weisman, J.; Ying, S. H.: A theoretically based critical heat flux prediction for rod bundles at PWR conditions. Nuclear Engineering and Design85 (1985) 239–25010.1016/0029-5493(85)90289-4Search in Google Scholar
11 Zhang, R.; Cong, T.; Tian, W.; et al.: Prediction of CHF in vertical heated tubes based on CFD methodology. Progress in Nuclear Energy78 (2015) 196–20010.1016/j.pnucene.2014.10.001Search in Google Scholar
12 AnsysI. ANSYS FLUENT theory guide. Canonsburg, PA, 2011: 794Search in Google Scholar
13 CD-adapco U. GUIDE: STAR-CCM+, 2009Search in Google Scholar
14 Ishii, M.; Zuber, N.: Drag coefficient and relative velocity in bubbly, droplet or particulate flows. AIChE Journal25 (1979) 843–85510.1002/aic.690250513Search in Google Scholar
15 Tomiyama, A.; Tamai, H.; Zun, I.; et al.: Transverse migration of single bubbles in simple shear flows. Chemical Engineering Science57 (2002) 1849–185810.1016/S0009-2509(02)00085-4Search in Google Scholar
16 Antal, S. P.; Lahey, R. T.; Flaherty, J. E.: Analysis of phase distribution in fully developed laminar bubbly two-phase flow. International Journal of Multiphase Flow17 (1991) 635–65210.1016/0301-9322(91)90029-3Search in Google Scholar
17 Yan, J.; Yuan, K.; Tatli, E.; et al.: CFD Prediction of Pressure Drop for the Inlet Region of a PWR Fuel Assembly. CFD4NRS-3 Hosted by United States Nuclear Regulatory Commission, Washington DC, USA, 2010Search in Google Scholar
18 dos Santos, A. A. Campagnole; Ribeiro, F. L.; Costa, F. C.; Filho, J. A. Barros; Navarro, M. A.: Verification and validation of a PWR rod bundle segment CFD simulation. The 15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, NURETH-15 NURETH15–389, Pisa, Italy, May 12–17, 2013Search in Google Scholar
19 Zhang, R.; Cong, T.; Tian, W.; et al.: Effects of turbulence models on forced convection subcooled boiling in vertical pipe. Annals of Nuclear Energy80 (2015) 293–30210.1016/j.anucene.2015.01.039Search in Google Scholar
20 Bartolomei, G.; Chanturiya, V.: Experimental study of true void fraction when boiling subcooled water in vertical tubes. Thermal Engineering14 (1967) 123–128Search in Google Scholar
21 Qi, S.; Xiaojun, W.; Zhao, X.; et al.: Experimental study of average void fraction in low-flow subcooled boiling. Atomic Energy Science and Technology, 39 (2005) 193–197.Search in Google Scholar
22 Han, B.; et al.: Numerical Study on Void Fraction Distribution of Subcooled Boiling in 5 × 5 Rod Bundle. Kunming. China: The Seventh China-Korea Workshop in Nuclear Reactor Thermal Hydraulics, 2015Search in Google Scholar
© 2016, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Challenges in reactor core thermal-hydraulics: subchannel analysis, CFD modeling and rod bundle CHF
- Technical Contributions/Fachbeiträge
- Subchannel analysis and correlation of the Rod Bundle Heat Transfer (RBHT) steam cooling experimental data
- CFD analysis on mixing effects of spacer grids with different dimples and sizes for advanced fuel assemblies
- An experimental investigation on dynamics and heat transfer associated with a single droplet impacting on a hot surface above the Leidenfrost point temperature
- Study on effects of mixing vane grids on coolant temperature distribution by subchannel analysis
- Reflood experiments in rod bundles with flow blockages due to clad ballooning
- The effect of spacer grid critical component on pressure drop under both single and two phase flow conditions
- Numerical method improvement for a subchannel code
- Numerical investigation on the characteristics of two-phase flow in fuel assemblies with spacer grid
- Effects of axial power shapes on CHF locations in a single tube and in rod bundle assemblies
- CFD evaluation on the thermohydraulic characteristics of tube support plates in steam generator
- Analysis of heat transfer under high heat flux nucleate boiling conditions
- Review of the correlation developments and a new concept based on mixing mechanism for heat transfer enhancement of spacer grids
- A comparison of the CFD simulation results in 5 × 5 sub-channels with mixing grids using different turbulence models
- Simulation of isothermal multi-phase fuel-coolant interaction using MPS method with GPU acceleration
- RELAP5 investigation on subchannel flow instability
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Challenges in reactor core thermal-hydraulics: subchannel analysis, CFD modeling and rod bundle CHF
- Technical Contributions/Fachbeiträge
- Subchannel analysis and correlation of the Rod Bundle Heat Transfer (RBHT) steam cooling experimental data
- CFD analysis on mixing effects of spacer grids with different dimples and sizes for advanced fuel assemblies
- An experimental investigation on dynamics and heat transfer associated with a single droplet impacting on a hot surface above the Leidenfrost point temperature
- Study on effects of mixing vane grids on coolant temperature distribution by subchannel analysis
- Reflood experiments in rod bundles with flow blockages due to clad ballooning
- The effect of spacer grid critical component on pressure drop under both single and two phase flow conditions
- Numerical method improvement for a subchannel code
- Numerical investigation on the characteristics of two-phase flow in fuel assemblies with spacer grid
- Effects of axial power shapes on CHF locations in a single tube and in rod bundle assemblies
- CFD evaluation on the thermohydraulic characteristics of tube support plates in steam generator
- Analysis of heat transfer under high heat flux nucleate boiling conditions
- Review of the correlation developments and a new concept based on mixing mechanism for heat transfer enhancement of spacer grids
- A comparison of the CFD simulation results in 5 × 5 sub-channels with mixing grids using different turbulence models
- Simulation of isothermal multi-phase fuel-coolant interaction using MPS method with GPU acceleration
- RELAP5 investigation on subchannel flow instability
