17 research outputs found
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Cylindrical shell buckling through strain hardening
Recently, the authors published results of plastic buckling analysis of cylindrical shells. Ideal elastic-plastic material behavior was used for the analysis. Subsequently, the buckling analysis program was continued with the realistic stress-strain relationship of a stainless steel alloy which does not exhibit a clear yield point. The plastic buckling analysis was carried out through the initial stages of strain hardening for various internal pressure values. The computer program BOSOR5 was used for this purpose. Results were compared with those obtained from the idealized elastic-plastic relationship using the offset stress level at 0.2% strain as the yield stress. For moderate hoop stress values, the realistic stress-grain case shows a slight reduction of the buckling strength. But, a substantial gain in the buckling strength is observed as the hoop stress approaches the yield strength. Most importantly, the shell retains a residual strength to carry a small amount of axial compressive load even when the hoop stress has exceeded the offset yield strength
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Age-Related Degradation of Nuclear Power Plant Structures and Components
This paper summarizes and highlights the results of the initial phase of a research project on the assessment of aged and degraded structures and components important to the safe operation of nuclear power plants (NPPs). A review of age-related degradation of structures and passive components at NPPs was performed. Instances of age-related degradation have been collected and reviewed. Data were collected from plant generated documents such as Licensing Event Reports, NRC generic communications, NUREGs and industry reports. Applicable cases of degradation occurrences were reviewed and then entered into a computerized database. The results obtained from the review of degradation occurrences are summarized and discussed. Various trending analyses were performed to identify which structures and components are most affected, whether degradation occurrences are worsening, and what was the most common aging mechanisms. The paper also discusses potential aging issues and degradation-susceptible structures and passive components which would have the greatest impact on plant risk
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Screening dynamic evaluation of SRS cooling water line
The production reactors at the Savannah River Site (SRS) have been shut down due to perceived safety concerns. A major concern is the seismic integrity of the plant. A comprehensive program is underway to assess the seismic capacity of the existing systems and components and to upgrade them to acceptable levels. The evaluation of the piping systems at the SRS is a major element of this program. Many of the piping systems at the production reactors were designed without performing dynamic analyses. Instead their design complied with good design practice for dead weight supported systems with proper accommodation of thermal expansion effects. In order to gain some insight as to the seismic capacity of piping installed in this fashion, dynamic analyses were performed for some lines. Since the piping was not seismically supported, the evaluations involved various approximations and the results are only used as a screening test of seismic adequacy. In this paper, the screening evaluations performed for the raw water inlet line are described. This line was selected for evaluation since it was considered typical of the smaller diameter piping systems at the plant. It is a dead weight supported system made up of a run of small diameter piping which extends for great distances over many dead weight supports and through wall penetrations. The results of several evaluations for the system using different approximations to represent the support system are described. 2 figs., 4 tabs
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Effect of PVRC damping with independent support motion response spectrum analysis of piping systems
The Technical Committee for Piping Systems of the Pressure Vessel Research Committee (PVRC) has recommended new damping values to be used in the seismic analyses of piping systems in nuclear power plants. To evaluate the effects of coupling these recommendations with the use of independent support motion analyses methods, two sets of seismic analyses have been carried out for several piping systems. One set based on the use of uniform damping as specified in Regulatory Guide 1.61, the other based on the PVRC recommendations. In each set the analyses were performed using independent support motion time history and response spectrum methods as well as the envelope spectrum method. In the independent response spectrum analyses, 14 response estimates were in fact obtained by considering different combination procedures between the support group contributions and all sequences of combinations between support groups, modes and directions. For each analysis set, the response spectrum results were compared with time history estimates of those results. Comparison tables were then prepared depicting the percentage by which the response spectrum estimates exceeded the time history estimates. By comparing the result tables between both analysis sets, the impact of PVRC damping can be observed. Preliminary results show that the degree of exceedance of the response spectrum estimates based on PVRC damping is less than that based on uniform damping for the same piping problem. Expressed differently the results obtained if ISM methods are coupled with PVRC damping are not as conservative as those obtained using uniform damping
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Nonlinear Finite-Element Analysis of a Reinforced-Concrete Mark III Containment Under Pressure and Gravity Loads.
An analysis of a Mark III reactor containment vessel subjected to a uniformly increasing internal pressure and gravity loads is carried out in order to ascertain the load carrying capacity of the structure under hydrogen burn. The analysis is conducted by using a nonlinear finite element model that includes nonlinearities in the strain-displacement relations as well as in the material constitutive equations. In this analysis, the nonlinear behavior of the liner and reinforcement steels is described by a von Mises elastic-plastic model with isotropic hardening. A recently developed elastic-plastic-fracture model that includes both the cracking and crushing limit states is used for the plain concrete. Consistent smearing and de-smearing procedures are then used to represent the composite material properties of the reinforced concrete by an anisotropic and locally homogeneous continuum. Results pertaining to the critical regions of the containment where cracking of the concrete, yielding of the reinforcement bars, and substantial straining of the liner take place are discussed
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BNL piping research
Brookhaven National Laboratory (BNL) has assisted in the development of methods to evaluate the analysis methods used by industry to qualify nuclear power piping. Through FY 1985 these efforts were conducted under the Mechanical Piping Benchmarks project while current and future efforts will be performed under the Combination Procedures for piping project. Under these projects BNL has developed analytical benchmark problems for piping systems evaluated using uniform or independent support motion response spectrum methods, investigated the adequacy and limitations of linear piping analysis methods by comparison to test results and evaluated and developed criteria for new and alternate methods of analysis. A summary description of the status of these efforts is provided
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Structural analysis of an underground reinforced concrete waste storage tank due to over-pressurization
This paper presents the results of a structural analysis performed by use of the finite element method in determining the pressure-carrying capacity of an underground tank which contains nuclear wastes. The tank and surrounding soil were modeled and analyzed using the ABAQUS program. Special emphases were given on determining the effects of soil-containment interaction by employing Coulomb friction model. The effect of material properties was investigated by considering two sets of stress-strain data for the steel plates. In addition, a refined mesh was used to evaluate the strain concentration effects at steel liner thickness discontinuities
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A report on the seismic capacity of the General Laboratory and Administration Building at Los Alamos National Laboratory
A seismic analysis of the General Laboratory and Administration Building at Los Alamos National Laboratory is performed. The analyses are performed in detail for one portion of the building and then qualitatively extrapolated to other portions of the building. Seismic capacities are evaluated based on two sets of acceptance criteria. The first is based on Code-type criteria and is associated with a low probability of failure. This capacity is found to be in the 0.04--0.06 G ZPA range (the free field seismic motion is defined with a NUREG 0098 response spectrum). The second capacity is based on much less conservative criteria such as might be associated with a high probability of failure. This capacity is found to be about 0.15 G. Finally structural modifications are proposed that would increase the low probability of failure capacity to 0.15 G ZPA. These modifications consist of steel double angle braces or concrete shear walls placed at some of the frames in the building
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Seismic fragility of nuclear power plant components (Phase 2): A fragility handbook on eighteen components
Fragility estimates of seven equipment classes were published in earlier reports. This report presents fragility analysis results from eleven additional equipment categories. The fragility levels are expressed in probabilistic terms. For users' convenience, this concluding report includes a summary of fragility results of all eighteen equipment classes. A set of conversion factors based on judgment is recommended for use of the information for early vintage equipment. The knowledge gained in conducting the Component Fragility Program and similar other programs is expected to provide a new direction for seismic verification and qualification of equipment. 15 refs., 12 tabs