A deterministic analysis of the IRIS safety features has been carried out by means of the best-estimate code RELAP (ver. RELAP5 mod3.2). First, the main system components were modeled and tested separately, namely: the Reactor Pressure Vessel (RPV), the modular helical-coil Steam Generators (SG) and the Passive (natural circulation) Emergency Heat Removal System (PEHRS). Then, a preliminary set of accident transients for the whole primary and safety systems was investigated. Since the project was in a conceptual phase, the reported analyses must be considered preliminary. In fact, neither the reactor components, nor the safety systems and the reactor signal logics were completely defined at that time. Three “conventional” design basis accidents have been preliminary evaluated: a Loss Of primary Flow Accident, a Loss Of Coolant Accident and a Loss Of Feed Water accident. The results show the effectiveness of the safety systems also in LOCA conditions; the core remains covered for the required grace period. This provides the basis to move forward to the preliminary design.
- Nuclear Engineering Division
Preliminary Safety Analysis of the IRIS Reactor
Ricotti, ME, Cammi, A, Cioncolini, A, Cipollaro, A, Oriolo, F, Lombardi, C, Conway, LE, & Barroso, AC. "Preliminary Safety Analysis of the IRIS Reactor." Proceedings of the 10th International Conference on Nuclear Engineering. 10th International Conference on Nuclear Engineering, Volume 2. Arlington, Virginia, USA. April 14–18, 2002. pp. 797-805. ASME. https://doi.org/10.1115/ICONE10-22398
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