CFD prediction of critical heat flux in vertical heated tubes with uniform and non-uniform heat flux
Creators
- 1. Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu 610213 (China)
- 2. Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)
- 3. Department of Engineering Physics, Tsinghua University, Beijing 100084 (China)
Description
Highlights: • Eulerian two-fluid model coupled with extended wall boiling model within a CFD code. • Simulation of DNB in vertical heated tubes under high pressures with STAR-CCM+ 10.04. • New methods developed to predict CHF in both uniform and non-uniform heated tubes. • The results showed some differences between uniform and non-uniform heated cases. - Abstract: In this paper, Eulerian Two-fluid model coupled with extended wall boiling model was used to simulate the departure from nucleate boiling (DNB) in vertical heated tubes under high pressures by using STAR-CCM+ 10.04. Based on CFD approach, new methods were developed to predict the critical heat flux (CHF) in both uniform and non-uniform heated tubes. The results showed some differences between the uniform and non-uniform heated cases. The transition in boiling curves was used as the criterion of DNB for uniform heat flux while the peaks of wall temperature and near wall void fraction were used for non-uniform cases. Good agreement was obtained between the predictions and experimental data, including both the values of critical heat flux and their locations. For uniform heat flux, sensitivity analysis for pressure, local quality and mass flow rate were performed, and the results showed that the new prediction method could work well under a wide range of conditions except under the high local quality and low mass flow rate conditions. The mechanism of this prediction method was discussed as well.
Availability note (English)
Available from http://dx.doi.org/10.1016/j.nucengdes.2017.11.009Additional details
Identifiers
- DOI
- 10.1016/j.nucengdes.2017.11.009;
- PII
- S0029549317305137;
Publishing Information
- Journal Title
- Nuclear Engineering and Design
- Journal Volume
- 326
- Journal Page Range
- p. 403-412
- ISSN
- 0029-5493
- CODEN
- NEDEAU
INIS
- Country of Publication
- Netherlands
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 50070958
- Subject category
- S42: ENGINEERING; S97: MATHEMATICAL METHODS AND COMPUTING;
- Resource subtype / Literary indicator
- Numerical Data
- Descriptors DEI
- CRITICAL HEAT FLUX; DEPARTURE NUCLEATE BOILING; EXPERIMENTAL DATA; FLOW RATE; FORECASTING; MATHEMATICAL MODELS; PRESSURE RANGE MEGA PA 10-100; SENSITIVITY ANALYSIS; SIMULATION; SUBCOOLED BOILING; TUBES; VOID FRACTION
- Descriptors DEC
- BOILING; DATA; HEAT FLUX; INFORMATION; NUCLEATE BOILING; NUMERICAL DATA; PHASE TRANSFORMATIONS; PRESSURE RANGE; PRESSURE RANGE MEGA PA
Optional Information
- Copyright
- Copyright (c) 2017 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.