Design, Instrumentation and Testing of a Steel Containment Vessel Model (NUREG/CR-5679)
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Publication Information
Manuscript Completed: February 1999
Date Published: February 2000
Prepared by:
V. K. Luk, M. F. Hessheimer, G. S. Rightley,
L. Dwight Lambert, E. W. Kiamerus
Sandia National Laboratories
Albuquerque, NM 87185-0744
J. F. Costello, NRC Project Manager
Prepared for:
Division of Engineering Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
Extreme Loads JCN A1401
Systems Safety Department
Nuclear Power Engineering Corporation
Tokyo, 105 Japan
Under Funds-in Agreement No. DE-F104-91AL73734
T. Hashimoto, NUPEC Project Manager
Availability Notice
Abstract
The Nuclear Power Engineering Corporation (NUPEC) of Japan and the US Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, are co-sponsoring and jointly funding a Cooperative Containment Research Program at Sandia National Laboratories. As a part of this program, a steel containment vessel model was tested to failure in the high pressure test on December 11-12, 1996. The model, which is representative of a steel containment for an improved Japanese Mark II Boiling Water Reactor Plant, has a geometric scale of 1:10 and a thickness scale of 1:4. The objectives of the steel containment vessel model test were to obtain measurement data of the structural response of the model up to its failure in order to validate analytical modeling, to find the pressure capacity of the model, and to observe the failure mechanisms.
The steel containment vessel model was surrounded by a contact structure, which provided a simplified representation of some features of actual concrete shield buildings in reactor plants. The model underwent radial expansion under internal pressurization, and the gap between the two structures closed in local areas where contact occurred. This special feature was designed to provide measurement data on load re-distribution on contact between the two structures to validate the contact algorithms in finite element codes.
The model and the contact structure were instrumented with strain gages, displacement transducers, contact detection devices, pressure transducers, and thermocouples. More than 97% of the installed instruments survived the high pressure test.
After 16.5 hours of continuous operation with monotonic increases in pressure, the high pressure test was terminated when a leakage was detected and the pressurization system at its maximum flow capacity could not maintain the pressure inside the model. The maximum pressure achieved during the high pressure test was 4.66 MPa (676 psig) or approximately 5.97 times the scaled design pressure. Posttest model inspection revealed that the leakage was caused by a large tear, approximately 190-mm-long, along a weld seam at the outside edge of the equipment hatch reinforcement plate. A small meridional tear, roughly 85-mm-long, was also discovered in a vertical weld inside a semi-circular weld relief opening at the middle stiffening ring above the equipment hatch.
The posttest metallurgical evaluation provided critical information on the deformation pattern and the failure mode and mechanisms of the two tears and on the strain concentrations in a few locally necked areas. All material deformation and tear observed in the samples, which were made from the sections removed from the model, were ductile in nature. There was no evidence of material flaws, defects, or brittle behavior in the base metal or welds. The tears that occurred resulted from exceeding the local plastic ductility of the alloy.
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