For severe accident assessment in a light water reactor (LWR), heat transfer models in a narrow annular gap between the overheated core debris and the reactor pressure vessel (RPV) are important for evaluating RPV integrity and emergency procedures. Some heat transfer models have been proposed as gap cooling CHF (critical heat flux) but local heat fluxes on the hot surface were not taken into account. Therefore, using the existing data, the authors developed heat transfer models on the average CHF restricted by CCFL (counter-current flow limitation) and local boiling heat fluxes, and showed that the average CHF depended on the steam-water flow pattern in the narrow gap and that the local heat fluxes were similar to the pool boiling curve. We evaluated the validity of heat transfer models by simple calculations for an ALPHA/IDC001 experiment performed by JAERI (Japan Atomic Energy Research Institute). Results showed heat fluxes on the crust surface were restricted mainly by its thermal resistance after the crust formation, emissivity on its surface did not have much effect on the heat fluxes, and the calculated vessel temperature during the heat-up process agreed well with the measurements. However, the vessel cooling rate was underestimated mainly due to underestimation of the gap size. The heat fluxes on the vessel inner surface were much higher than the pool film boiling therefore local boiling heat transfer should be studied to improve the heat transfer models.
- Nuclear Engineering Division
Evaluation of Debris Cooling Models in RPV Lower Head Based on Analysis for JAERI-ALPHA Test
Murase, M, Kohriyama, T, Yoshida, Y, & Okano, Y. "Evaluation of Debris Cooling Models in RPV Lower Head Based on Analysis for JAERI-ALPHA Test." Proceedings of the 10th International Conference on Nuclear Engineering. 10th International Conference on Nuclear Engineering, Volume 3. Arlington, Virginia, USA. April 14–18, 2002. pp. 95-103. ASME. https://doi.org/10.1115/ICONE10-22098
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