Classical problems of columns under tangential (follower) loads are analyzed for stability. A generalized Ritz technique is used in conjunction with a computer-oriented algorithm suitable for the flutter analysis of complex structural systems. A brief description of this technique is presented. The numerical approach allows for easy consideration of additional effects in classical problems such as the Beck and Leipholz columns. The influence of shear deformation, rotatory inertia, and concentrated mass at the tip of the column is shown in graphs and discussed.

During the 8th outage of the Angra 2 power plant, the expansion joints of the containment penetration JMK05D0201 (LCQ50) were found distorted. The cause was an inadvertent over pressurization made during monthly tests to verify the set pressure. To leave the damaged expansion joint installed, Eletronuclear conducted an elastic-plastic analysis which has demonstrated its integrity. Eletronuclear decided to keep it in place and cover it with an expansion joint designed to resist all loads predicted in the original design specification. This new design is a two-sided (bipartite) bellows with longitudinal field welds. This chapter presents the fatigue and stress analysis of the bipartite overlapping expansion joints considering the loads defined in the design specification.

Qualified fatigue assessment based on realistic input data constitutes an essential part of an ageing management strategy for Nuclear Power Plants. In this context and as a continuation of a previous paper PVP2014-28716 the requirements of load data evaluation, stress analysis and cycle counting are detailed based on a real world example from a Brazilian Nuclear Power Plant. One essential prerequisite of any fatigue assessment approach is the availability of realistic load data. In the present analysis, selected operational plant data from the period 2003 to 2012 are used. One further prerequisite is the accurate component stress analysis based on a transient thermal-mechanical Finite Element Analyses. As an example, a highly loaded nozzle from the Chemical & Volume Control System (CVCS) is chosen to be analyzed. The influences on the fatigue assessment caused by the load-time histories, the stress analysis approaches and the cycle counting method are discussed in detail. The considered operational time period from 2003 to 2012 with respective selected plant data gives a consolidated background. It is one essential aim of the study to show the influence of the load-data input and the (design code conforming) stress analysis method on the resulting calculated cumulative usage factors (CUFs). In the present paper, the stress analysis employs the finite element method. Simplified elastic-plastic (application of ke plasticity factors) procedures are used in order to identify the margins and influences of design and actual loading histories on the resulting CUFs. The paper concludes with a comprehensive picture including quantification and discussion of the different influencing parameters on the resulting CUFs. This reveals margins in the fatigue design process and solutions of coping with the design code requirements.

The penetrations in the early Pressurized Water Reactors Vessels are characterized by Alloy 600 tubes, welded by Alloy 182/82. The Alloy 600 tubes have been shown to be susceptible to PWSCC (Primary Water Stress Corrosion Cracking) which may lead to crack forming. The cracking mechanism is driven mainly by the welding residual stress and, in a second place, by the operational stress in the weld region. It is therefore of big interest to quantify the weld residual stress field correctly. In this paper the weld residual stress field is calculated by finite elements, using a common approach well known in nuclear domain. It includes a transient thermal analysis simulating the heating during the multipass welding, followed by a transient thermo-mechanical analysis for the determination of the stresses involved with it. The welding consists of a sequence of weld beads, each of which is deposited in its entirety, at once, instead of gradually. Central as well as eccentric sidehill nozzles on the vessel head are analyzed in the paper. For the former a 2-dimensional axisymmetrical finite element model is used, whereas for the latter a 3-dimensional model is set up. Different positions on the vessel head are compared and the influence of the sidehill effect is illustrated. In the framework of a common project for Angra 1, Tractebel Engineering (Belgium) and Eletronuclear (Nuclear Utility, Brazil) had the opportunity to compare their analysis method, which they applied to the Belgian and the Brazilian nuclear reactors, respectively. The global approach in both cases is very similar but is applied to different configurations, specific for each NPP. In the article the results of both cases are compared.

The steam generator (SG) snubber elimination process in a nuclear power plant requires a new evaluation of the structural behavior of the complete primary system components for licensing purposes. The forces and stresses have to be evaluated in all supports, piping, nozzles and internals of all components of the reactor coolant loop (RCL) for the required load cases, including dead weight, thermal conditions, seismic excitations and postulated piping ruptures. The SG snubber elimination intends to obtain a safer operating condition, avoiding problems with snubber maintenance, inspection and mal-function. The paper describes the methodology adopted for this type of analysis, where a very detailed modeling procedure is required, both for the primary loop itself, where nonlinearities are introduced to represent the supporting devices, as well as for the coupling with the reactor building structure. The piping and the components (Reactor vessel, SG and pumps) are modeled in order to represent their weight distribution, stiffness and supporting conditions in detail. The reactor building complete 3D-finite element model is reduced to a corresponding representative simple beam model in order to make the nonlinear dynamic analysis feasible. The seismic response spectra from both building models were compared at supporting points of the primary circuit in order to guarantee that the simple beam model represents the behavior of the refined building model in a correct way. The dynamic analysis is performed with seismic acceleration time histories applied at the foundation of the reactor building model and a direct integration method is used. The Rayleigh damping values as well as the effects in the results of refining the integration time steps are discussed. The impact forces due to postulated pipe ruptures are also evaluated as impact loads. The results of these analyses are displacements, accelerations and forces in all structural elements and their supports, as well as time histories and response spectra for the stress analysis of the component internal structures.