Abstract
The prediction of the peak cladding temperature is key for the evaluation of the safety characteristics of nuclear reactor designs. One of the most important parameters in determining the peak cladding temperature is the two-phase mixture level in the core, which in turn depends on the void fraction in the core. In the two-fluid model, which is used in best-estimate thermal-hydraulic analysis codes such as RELAP and TRACE, the void fraction is largely determined by the interfacial drag. In the codes, the interfacial drag is determined based on drift-flux type correlations for the distribution parameter and void-weighted area-averaged drift velocity. It is essential that the model for the prediction of void fraction in the reactor core be accurate to ensure safe shutdown of the reactor. In light of this, a thorough review of the Chexal-Lellouche drift-flux type correlation currently used in RELAP5/MOD3.2.2 has been initiated to evaluate the scalability and uncertainty of the correlation. Analysis of the physical dependencies indicates some serious concerns regarding the presence of compensating errors in the models for the distribution parameter and void-weighted area-average drift velocity. © 2014 Elsevier Ltd. All rights reserved.
| Original language | English |
|---|---|
| Pages (from-to) | 143-153 |
| Journal | Progress in Nuclear Energy |
| Volume | 74 |
| DOIs | |
| Publication status | Published - Jul 2014 |
| Externally published | Yes |
Bibliographical note
Publication details (e.g. title, author(s), publication statuses and dates) are captured on an “AS IS” and “AS AVAILABLE” basis at the time of record harvesting from the data source. Suggestions for further amendments or supplementary information can be sent to [email protected].Research Keywords
- CSAU
- Drift-flux model
- Rod bundle
- Uncertainty
- Void fraction