Publications

Preloads are routinely applied to stiffen structural members in many applications. However, the preloaded structural members have been observed to lose a significant portion of the imposed load due to internal relaxation mechanisms during impulsive impact events. This paper describes the design and initial experiments for a novel Hopkinson bar configuration designed to investigate the effect of preloads on the stress wave propagation across interfaces between the incident and transmission bars. Dynamic responses are measured by a variety of sensors, including accelerometers, strain gages, and a laser vibrometer. The transmissibility of a titanium incident bar is measured to establish the baseline frequency response between the input and the test interface. Wave transmission across an titanium-aluminum interface is also examined by analyzing the frequency response function, transmission efficiency, and transmissibility between the incident and transmitted waves. The presence of vacuum grease is shown to strongly influence the dynamic behavior of the system.

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Determination of fracture toughness for metals under quasi-static loading conditions can follow well-established procedures and ASTM standards. The use of metallic materials in impact-related applications requires the determination of dynamic fracture toughness for these materials. There are two main challenges in experiment design that must be overcome before valid dynamic data can be obtained. Dynamic equilibrium over the entire specimen needs to be approximately achieved to relate the crack tip loading state to the far-field loading conditions. The loading rate at the crack tip should be maintained nearly constant during an experiment to delineate rate effects on the values of dynamic fracture toughness. A recently developed experimental technique for determining dynamic fracture toughness of brittle materials has been adapted to measure the dynamic initiation fracture toughness of high strength steel alloys. A split-Hopkinson pressure bar is used to apply the dynamic loading. A pulse shaper is used to achieve constant loading rate at the crack tip and dynamic equilibrium across the specimen. A four-point bending configuration is used at the impact section of the setup. ©2008 Society for Experimental Mechanics Inc.

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The overall goal of this joint research project was to develop and demonstrate advanced sensors and computational technology for continuous monitoring of the condition of components, structures, and systems in advanced and next-generation nuclear power plants (NPPs). This project included investigating and adapting several advanced sensor technologies from Korean and US national laboratory research communities, some of which were developed and applied in non-nuclear industries. The project team investigated and developed sophisticated signal processing, noise reduction, and pattern recognition techniques and algorithms. The researchers installed sensors and conducted condition monitoring tests on two test loops, a check valve (an active component) and a piping elbow (a passive component), to demonstrate the feasibility of using advanced sensors and computational technology to achieve the project goal. Acoustic emission (AE) devices, optical fiber sensors, accelerometers, and ultrasonic transducers (UTs) were used to detect mechanical vibratory response of check valve and piping elbow in normal and degraded configurations. Chemical sensors were also installed to monitor the water chemistry in the piping elbow test loop. Analysis results of processed sensor data indicate that it is feasible to differentiate between the normal and degraded (with selected degradation mechanisms) configurations of these two components from the acquired sensor signals, but it is questionable that these methods can reliably identify the level and type of degradation. Additional research and development efforts are needed to refine the differentiation techniques and to reduce the level of uncertainties.

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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.

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A high pressure test of the steel containment vessel (SCV) model was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. A concentric steel contact structure (CS), installed over the SCV model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. The SCV model and contact structure were instrumented with strain gages and displacement transducers to record the deformation behavior of the SCV model during the high pressure test. This paper summarizes the conduct and the results of the high pressure test and discusses the posttest metallurgical evaluation results on specimens removed from the SCV model.

A high pressure test of a scale model of a steel containment vessel (SCV) was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. This testis part of a program to investigate the response of representative models of nuclear containment structures to pressure loads beyond the design basis accident. The posttest analyses of this test focused on three areas where the pretest analysis effort did not adequately predict the model behavior during the test. These areas are the onset of global yielding, the strain concentrations around the equipment hatch and the strain concentrations that led to a small tear near a weld relief opening that was not modeled in the pretest analysis.

A high pressure test of the steel containment vessel (SCV) model was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. Several organizations from the US, Europe, and Asia were invited to participate in a Round Robin analysis to perform independent pretest predictions and posttest evaluations of the behavior of the SCV model during the high pressure test. Both pretest and posttest analysis results from all Round Robin participants were compared to the high pressure test data. This paper summarizes the Round Robin analysis activities and discusses the lessons learned from the collective effort.

This report discusses fourteen tests which were conducted to investigate the perforation of thin unreinforced concrete slabs. The 4340-steel projectile used in the test series is 50.8 mm in diameter, 355.6 mm in length, has a mass of 2.34 kg. and an ogive nose with caliber radius head of 3. The slabs, contained within steel culverts, are 1.52 m in diameter and consist of concrete with a nominal unconfined compressive strength of 38.2 MPa and maxima aggregate size of 9.5 mm. Slab thicknesses are 284.4, 254.0, 215.9 and 127.0 mm. Tests were conducted at impact velocities of about 313 m/s on all slab thicknesses and about 379 and 471 m/s on the 254.0-mm-thick slab. All tests were conducted at normal incidence to the slab. All tests were conducted at normal incidence to the slab. Information obtained from the tests used to determine the loading (deceleration) on the projectile during the perforation process, the velocity-displacement of the projectile as it perforated the slab, and the projectile position as damage occurred on the backface of the slab. The test projectile behaved essentially as a rigid body for all of the tests.