The ASTRID (Assessment of Source Term for Emergency Response based on Installation Data) process model is used for the faster than real-time prediction of the radioactivity released into the containment and further into the environment in case of an emergency situation in a light water reactor. Combined together with the containment module COCOSYS the model can predict the entire radioactivity release chain from the primary system to the containment and further into the environment. In the paper the ASTRID thermohydraulic module PROCESS is presented shortly. The thermohydraulic part is a fast running solution for the drift-flux based thermohydraulics. In high temperatures the core degradation leading to the melt pool formation in the reactor barrel and reactor vessel lower head is calculated in the in-vessel module RELOMEL. Finally after the reactor vessel wall has been eroded due to the molten corium in the lower plenum, the massive radioactivity release occurs into the containment. But even before this scenario the radioactivity may be transported from the superheated core to the containment by the coolant. The reference plants for the development have been the Westinghouse type 4-loop PWR, the French type 3-loop PWR, The German type 4-loop Konvoi PWR, the Loviisa VVER type PWR, and the Olkiluoto type internal pump BWR. The reference code for the DBA thermal hydraulics has been the SMABRE code. In the developmental assessment the capability of the rough nodalization of ASTRID has been tested against the SMABRE nodalization describing the plants with 50–500 nodes. For the developmental assessment of the in-vessel severe accident the sample cases are calculated with MELCOR. The more thorough validation is based on the internationally known system codes, RELAP5, MELCOR, CATHARE and ATHLET. In the validation the most problematic area is the radioactivity transport into the containment. This part of the validation is done with the integrated code system.
- Nuclear Engineering Division
Validation of the Fast-Running In-Vessel Model ASTRID for Predicting the Radioactive Releases to the Containment
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Jaakko, M, & Schmuck, P. "Validation of the Fast-Running In-Vessel Model ASTRID for Predicting the Radioactive Releases to the Containment." Proceedings of the 12th International Conference on Nuclear Engineering. 12th International Conference on Nuclear Engineering, Volume 3. Arlington, Virginia, USA. April 25–29, 2004. pp. 211-221. ASME. https://doi.org/10.1115/ICONE12-49454
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