Numerical simulation of the effect of rod bowing on critical heat flux
-
Bing Ren
,
Shiyin Xu
Abstract
The paper demonstrates the CFD capability to predict the effect of rod bowing on CHF. The applicability and performance of the models are validated based on the Weatherhead experimental data for vertical pipe and NUPEC PWR subchannel and bundle test (PSBT) International Benchmark for rod bundle. The capability of the existing CFD method in predicting CHF is studied both qualitatively and quantitatively. Based on the validated method, the effect of the bowed rod on CHF is evaluated. The bow is assumed to occur at the midspan between two spacer grids in the region where DNB is predicted to occur. The results indicate that rod bowing has a deleterious effect on CHF, but this adverse bowing effect is dependent on the closure (i.e., the ratio of the change of the fuel rod gap to the nominal gap). There is no effect on CHF for closure between 0 and 50%, but adverse effect on CHF for closure larger than 50%, even more considerable effect as the closure increases to larger than 85%. The findings can be used to assist in designing the test fuel rod before the rod bowing CHF experiment.
Acknowledgement
The support provided by OECD/NEA and JNES is gratefully acknowledged. The data used in this study is from the OECD/PSBT benchmark of OECD/NEA and JNES.
-
Author contributions: All the authors have accepted responsibility for the entire content of this submitted manuscript and approved submission.
-
Research funding: None declared.
-
Conflict of interest statement: The authors declare no conflicts of interest regarding this article.
References
ANSYS 12.0 (2014). ANSYS Inc., Canonsburg PA, USA.Search in Google Scholar
Brewster, R.A., Baglietto, E., and Pointer, W.D. (2018). Assessment of boiling models for CFD-based prediction of DNB on Westinghouse PWR fuel grids. CFD4NRS-7, Shanghai.Search in Google Scholar
Burns, A.D., Frank, T., and Hamill, I. (2004). The Favre averaged drag model for turbulent dispersion in Eulerian multi-phase flows. Proceedings of the 5th International Conference on Multiphase Flow, Yokohama, 2004.Search in Google Scholar
Cole, R.A. (1960). Photographic study of pool boiling in the region of the critical heat flux. AIChE J. 533–538, https://doi.org/10.1002/aic.690060405.Search in Google Scholar
Kelly, J.E., Stewart, S.W., and Cuta, J.M. (1992). VIPRE-02: a two-fluid thermal-hydraulics code for reactor core and vessel analysis: mathematical modeling and solution methods. Nucl. Technol. 100: 246–259, https://doi.org/10.13182/NT92-A34746.Search in Google Scholar
Kurul, N. and Podowski, M.Z. (1990). Multidimensional effects in forced convection subcooled boiling. Proceedings of the 9th International Heat Transfer Conference, Israel.10.1615/IHTC9.40Search in Google Scholar
Li, Q., Avramova, M., and Jiao, Y.J. (2018). CFD prediction of critical heat flux in vertical heated tubes with uniform and non-uniform heat flux. Nucl. Eng. Des. 326: 403–412, https://doi.org/10.1016/j.nucengdes.2017.11.009.Search in Google Scholar
Lemmert, M. and Chawla, J.M. (1977). Influence of flow velocity on surface boiling heat transfer coefficient. Heat Transfer in boiling. Academic Press and Hemisphere, Cambridge.Search in Google Scholar
Mimouni, S., Baudry, C., and Guingo, M. (2016). Computational multi-fluid dynamics predictions of critical heat flux in boiling flow. Nucl. Eng. Des. 299: 28–36, https://doi.org/10.1016/j.nucengdes.2015.07.017.Search in Google Scholar
Morel, C. (2015). Mathematical modeling of disperse two-phase flows. Springer, Berlin.10.1007/978-3-319-20104-7Search in Google Scholar
Nagino, Y., Shodai, N., Matsumoto, C., and Masao, Ueji (1978). Rod bowed to contact departure from nucleate tests in cold wall thimble cell geometry. J. Nucl. Sci. Technol. 15: 568–573. https://doi.org/10.1080/18811248.1978.9735551.Search in Google Scholar
Nakajima, I., Kikuchi, A., and Kobori, T. (1977). A study of the effect of rod bowing on critical heat flux. Nucl. Eng. Des. 42: 237–245, https://doi.org/10.1016/0029-5493(77)90185-6.Search in Google Scholar
Pothukuchi, H., Kelm, S., and Patnaik, B.S.V. (2019). CFD modeling of critical heat flux in flow boiling: validation and assessment of closure models. Appl. Therm. Eng. 150: 651–665, https://doi.org/10.1016/j.applthermaleng.2019.01.030.Search in Google Scholar
Ranz, W.R. (1952). Evaporation from drops Part I & II. Chem. Eng. Prog. 48: 141–173.Search in Google Scholar
Rubin, A., Schoedel, A., and Avramova, M. (2010). OECD/NRC benchmark based on NUPEC PWR subchannel and bundle tests(PSBT). Volume I: experimental database and final problem specifications report NEA/NSC/DOC(2012)1.US NRC. OECD Nuclear Energy Agency, Paris.10.1155/2013/946173Search in Google Scholar
Siemens NX 12.0 (2017). Siemens product lifecycle management software Inc.Siemens, Germany.Search in Google Scholar
Tolubinsky, V.I. and Kostanchuk, D.M. (1970). Vapor bubbles growth rate and heat transfer intensity at subcooled water boiling. Proceedings of the 4th International Heat Transfer Conference, Paris.10.1615/IHTC4.250Search in Google Scholar
Tomiyama, A., Kataoka, I., Zun, I., and Sakaguchi, T. (1998). Drag coefficients of single bubbles under normal and micro gravity conditions. JSME Int. J.-Ser. B Fluids Therm. Eng. 41: 472–479, https://doi.org/10.1299/JSMEB.41.472.Search in Google Scholar
Weatherhead, R.J. (1963). Nuclear boiling characteristics and the critical heat flux occurrence in sub-cooled axial flow water systems. Argonne National Laboratory, Illinois.10.2172/4727562Search in Google Scholar
Zhang, R., Duarte, J.P., Cong, T.L., and Corradini, M. (2019). Investigation on the critical heat flux in a 2 by 2 fuel assembly under low flow rate and high pressure with a CFD methodology. Ann. Nucl. Energy 124: 69–79, https://doi.org/10.1016/j.anucene.2018.09.033.Search in Google Scholar
© 2021 Walter de Gruyter GmbH, Berlin/Boston
Articles in the same Issue
- Frontmatter
- Single-phase flow heat transfer characteristics in helically coiled tube heat exchangers
- Design and optimization of an integrated gamma ray scanning system for the uranium sample
- Numerical simulation of the effect of rod bowing on critical heat flux
- Flow and heat transfer characteristics of a nanofluid as the coolant in a typical MTR core
- Mathematical modeling of point kinetic equations with temperature feedback for reactivity transient analysis in MTR
- An enhanced formalism for the inverse reactor kinetics problem
- The establishment of analysis methodology of NRCDose3 for Kuosheng nuclear power plant decommissioning
- Analysis of SMART reactor core with uranium mononitride for prolonged fuel cycle using OpenMC
- Conceptual design of an innovative I&XC fuel assembly for a SMR based on neutronic/thermal-hydraulic calculations at the BOC
- Optimized fractional-order PID controller based on nonlinear point kinetic model for VVER-1000 reactor
- High rate X-ray radiation shielding ability of cement-based composites incorporating strontium sulfate (SrSO4) minerals
- Vibration analysis for predictive maintenance and improved reliability of rotating machines in ETRR-2 research reactor
- Calendar of events
Articles in the same Issue
- Frontmatter
- Single-phase flow heat transfer characteristics in helically coiled tube heat exchangers
- Design and optimization of an integrated gamma ray scanning system for the uranium sample
- Numerical simulation of the effect of rod bowing on critical heat flux
- Flow and heat transfer characteristics of a nanofluid as the coolant in a typical MTR core
- Mathematical modeling of point kinetic equations with temperature feedback for reactivity transient analysis in MTR
- An enhanced formalism for the inverse reactor kinetics problem
- The establishment of analysis methodology of NRCDose3 for Kuosheng nuclear power plant decommissioning
- Analysis of SMART reactor core with uranium mononitride for prolonged fuel cycle using OpenMC
- Conceptual design of an innovative I&XC fuel assembly for a SMR based on neutronic/thermal-hydraulic calculations at the BOC
- Optimized fractional-order PID controller based on nonlinear point kinetic model for VVER-1000 reactor
- High rate X-ray radiation shielding ability of cement-based composites incorporating strontium sulfate (SrSO4) minerals
- Vibration analysis for predictive maintenance and improved reliability of rotating machines in ETRR-2 research reactor
- Calendar of events
