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A 1:4-scale model of a prestressed concrete containment vessel (PCCV), representative of a pressurized water reactor (PWR) plant in Japan, was constructed by NUPEC at Sandia National Laboratories from January 1997 through June, 2000. Concurrently, Sandia instrumented the model with nearly 1500 transducers to measure strain, displacement and forces in the model from prestressing through the pressure testing. The limit state test of the PCCV model, culminating in functional failure (i.e. leakage by cracking and liner tearing) was conducted in September, 2000 at Sandia National Laboratories. After inspecting the model and the data after the limit state test, it became clear that, other than liner tearing and leakage, structural damage was limited to concrete cracking and the overall structural response (displacements, rebar and tendon strains, etc.) was only slightly beyond yield. (Global hoop strains at the mid-height of the cylinder only reached 0.4%, approximately twice the yield strain in steel.) In order to provide additional structural response data, for comparison with inelastic response conditions, the PCCV model filled nearly full with water and pressurized to 3.6 times the design pressure, when a catastrophic rupture occurred preceded only briefly by successive tensile failure of several hoop tendons. This paper summarizes the results of these tests.

The purpose of the program is to investigate the response of representative scale models of nuclear containment to pressure loading beyond the design basis accident and to compare analytical predictions to measured behavior. This objective is accomplished by conducting static, pneumatic overpressurization tests of scale models at ambient temperature. This research program consists of testing two scale models: a steel containment vessel (SCV) model (tested in 1996) and a prestressed concrete containment vessel (PCCV) model, which is the subject of this paper.

The Nuclear Power Engineering Corporation of Japan and the US Nuclear Regulatory Commission Office of Nuclear Regulatory Research, are co-sponsoring and jointly funding a Cooperative Containment Research Program at Sandia National Laboratories, Albuquerque, New Mexico, USA. As a part of this program, a steel containment vessel model and contact structure assembly was tested to failure with over pressurization at Sandia on December 11--12, 1996. The steel containment vessel model was a mixed-scale model (1:10 in geometry and 1:4 in shell thickness) of a steel containment for an improved Mark-II Boiling Water Reactor plant in Japan. The contact structure, which is a thick, bell-shaped steel shell separated at a nominally uniform distance from the model, provides a simplified representation of features of the concrete reactor shield building in the actual plant. The objective of the internal pressurization test was to provide measurement data of the structural response of the model up to its failure in order to validate analytical modeling, to find its pressure capacity, and to observe the failure model and mechanisms.