[en] The authors investigated the effects of torus wall flexibility in the pressure suppression system of a Mark I boiling water reactor (BWR) when the torus wall is subjected to hydrodynamic loadings. Using hypothetical models, they examined these flexibility effects under two hydrodynamic loading conditions: (1) a steam relief valve (SRV) discharge pulse, and (2) a loss-of-coolant accident (LOCA) chugging pulse. In the analyses of these events they used a recently developed two-dimensional finite element computer code. Taking the basic geometry and dimensions of the Monticello Mark I BWR nuclear power plant (in Monticello, Minnesota, U.S.A.), they assessed the effects of flexibility in the torus wall by changing values of the inside-diameter-to-wall-thickness ratio. Varying the torus wall thickness (t) with respect to the inside diameter (D) of the torus, they assigned values to the ratio D/t ranging from 0 (infinitely rigid) to 600 (highly flexible). In the case of a modeled steam relief valve (SRV) discharge pulse, they found the peak vertical reaction force on the torus was reduced from that of a rigid wall response by a factor of 3 for the most highly flexible, plant-simulated wall (D/t = 600). The reduction factor for a modeled loss-of-coolant accident (LOCA) chugging pulse was shown to be 1.5. The two-dimensional analyses employed overestimate these reduction factors but have provided, as intended, definition of the effect of torus boundary stiffness. In the work planned for FY79, improved modeling of the structure and of the source is expected to result in factors more directly applicable to actual pressure suppression systems