This thesis presents a model for predicting the forced convective critical heat flux (CHF) for water over a wide range of thermal-hydraulic conditions which might be encountered during normal and accident operations of a light water nuclear reactor. The model is primarily composed from existing steady-state CHF correlations for tubes or tube and rod bundle geometries, and encompasses the following parametric ranges:
0.3 ≤ P (MPa) ≤ 16.0
6.0 ≤ D (mm) ≤ 30.0
100.0 ≤ G (kg/m2s) ≤ 8000.0
-0.30 ≤ X ≤ 1.0
The correlations used as the foundation of this model are the
1) Westinghouse-3
2) Biasi correlation, and the
3) Modified Barnett correlation
The mode 1 presented is comp a red with available data, and the resultant model is illustrated as a 3-D surface in mass flux, quality, and CHF space to represent general CHF behavior.
| Identifer | oai:union.ndltd.org:ucf.edu/oai:stars.library.ucf.edu:rtd-1674 |
| Date | 01 January 1983 |
| Creators | Dahlquist, Joseph E. |
| Publisher | University of Central Florida |
| Source Sets | University of Central Florida |
| Language | English |
| Detected Language | English |
| Type | text |
| Format | application/pdf |
| Source | Retrospective Theses and Dissertations |
| Rights | Public Domain |
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