Experiments to determine the leakage behavior of pressure-unseating equipment hatches

Under the sponsorship of the United States Nuclear Regulatory Commission, Sandia National Laboratories is conducting several research programs to develop methods for estimating the ultimate pressure capacity, at elevated temperatures, of light water reactor containment buildings. These programs are collectively known as the Containment Integrity Programs. The developed methods will be used to assess the performance of containments in the highly unlikely event of a sever accident. A series of thirteen tests has been conducted to determine the leakage behavior of pressure-unseating equipment hatches. The tested unseating equipment hatch is a part of the 1/6-scale reinforced concrete containment model at Sandia. During the test series, the effect of the following parameters on leakage was observed: bolt preload and stiffness, gasket material, aging, and temperature. A summary of the findings of this test program is provided below: (1) increasing bolt preload and/or stiffness effectively delayed leakage to higher pressure levels; (2) leakage was not significantly affected by gasket material, aging, or test temperature up to the gasket material degradation temperature, and (3) there is considerable margin between the leakage onset and unseating pressures up to the gasket material degradation temperature. 6 refs., 3 figs., 2 tabs.