An analysis of design basis accidents has been completed and an analysis of the consequences of postulated beyond design basis accidents is now being performed on the IR-8 reactor.A computational analysis of the consequences of a beyond design basis accident with rupture of the delivery and intake piping in the first cooling loop of the reactor is being done. Preliminary assessments were initially made using the onedimensional code BEREZA. In order to confirm the results and determine the time to onset of core drying, a three-dimensional thermohydraulic model of IR-8 was developed and constructed, and a computational analysis of this accident was performed using the systems code ATHLET.At the National Research Center Kurchatov Institute, an analysis of design basis accidents has been completed and an analysis of beyond design basis accidents on IR-8 upon switching from high-enrichment fuel (90% 235 U) to low-enrichment fuel (19.7% 235 U) is being performed. An analysis of the consequences of the postulated beyond design basis accidents must be performed for the following cases: 1) complete instantaneous pipe rupture in the first cooling loop of the reactor; 2) complete instantaneous rupture of a horizontal experimental channel (damage to the horizontal channel); 3) complete instantaneous dewatering of the storage facility for spent fuel assemblies; and 4) spontaneous extraction of an automatic control rod followed by loss of forced circulation upon failure of both emergency protection rods.Design basis accidents due to insertion of positive reactivity were analyzed using the PARET code (ANL, USA), recommended by IAEA for calculations of research reactors [1]. The PARET code analyzes the development and consequences of reactivity and loss-of-flow accidents. This code uses a hydrodynamic model and a point kinetics model to calculate transient and emergency processes in a reactor. Except, perhaps, for the last case, viz., spontaneous extraction of an automatic control rod, this code cannot be used to analyze beyond design basis accidents on IR-8.An analysis of the consequences of instantaneous pipe rupture in the first cooling loop of the reactor has been completed. The preliminary assessments were initially performed using the one-dimensional code BEREZA [2]. In order to confirm the results and determine the time to onset of core dewatering, a three-dimensional thermohydraulic model of IR-8 was
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