Results 1 - 10 of 89
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[en] The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)
[en] Among the objectives marked by the Generation IV International Forum (GIF) is to provide the new generation of nuclear reactors of a level of safety equal to or greater than the current generation. Different European initiatives (SNTP, ESNII) have established r and d efforts to attain these ambitious objectives on its agenda. Among them is the development of tools for simulating the behavior of these designs in terms of accident in a precise and detailed. This article describes the different succession stages to develop a preliminary design of the European Sodium Fast Reactor (ESFR) model. On the basis of the development TRACE of a one-dimensional with neutron kinetics strut feedback model contrasted with the results obtained with equivalent models was taken as a reference, was extended to a three dimensional model-coded thermohydraulic to later be coupled with a neutron's kinetic space (PARCS) code. The conclusion of the work is the demonstration that conventional calculation tools can be adapted and used in the analysis of advanced reactor safety. Also identifies applications and developments of the coupled model and its implications in the study of safety of the reactor. (Author)
[en] Surveillance requirements and maintenance activities in a nuclear power plant aim to preserve components' inherent reliability. Up to now, predictive and preventive maintenance mainly concerned plant staff, but the US Nuclear Regulatory Commission Maintenance Rule released in July 1991 will have significant impact on how nuclear power plants perform and document this maintenance. Reliability Centered Maintenance (RCM) is a systematic methodology to establish maintenance tasks for critical components in plant with a high degree of compliance with the goals of the Rule. RCM pursues the identification of applicable and efficient tasks to prevent these components from developing their dominant failure causes, and, in turn, towards achieving proper levels of components availability with low cost. In this paper, we present an approach for identifying the most suitable set of tasks to achieve this goal, which involves the integration of maintenance activities and surveillance requirements for each critical component based on the unavailability and cost associated with each individual task which is performed on it
[en] There is a growing interest from both the regulatory authorities and the nuclear industry to stimulate the use of Probabilistic Risk Analysis (PRA) for risk-informed applications at Nuclear Power Plants (NPPs). Nowadays, special attention is being paid on analyzing plant-specific changes to Test Intervals (TIs) within the Technical Specifications (TSs) of NPPs and it seems to be a consensus on the need of making these requirements more risk-effective and less costly. Resource versus risk-control effectiveness principles formally enters in optimization problems. This paper presents an approach for using the PRA models in conducting the constrained optimization of TIs based on a steady-state genetic algorithm (SSGA) where the cost or the burden is to be minimized while the risk or performance is constrained to be at a given level, or vice versa. The paper encompasses first with the problem formulation, where the objective function and constraints that apply in the constrained optimization of TIs based on risk and cost models at system level are derived. Next, the foundation of the optimizer is given, which is derived by customizing a SSGA in order to allow optimizing TIs under constraints. Also, a case study is performed using this approach, which shows the benefits of adopting both PRA models and genetic algorithms, in particular for the constrained optimization of TIs, although it is also expected a great benefit of using this approach to solve other engineering optimization problems. However, care must be taken in using genetic algorithms in constrained optimization problems as it is concluded in this paper
[en] Lead-Cooled Fast Reactor (LFR) has been identified as one of promising future reactor concepts in the technology road map of the Generation IVC International Forum (GIF)as well as in the Deployment Strategy of the European Sustainable Nuclear Industrial Initiative (ESNII), both aiming at improved sustainability, enhanced safety, economic competitiveness, and proliferation resistance. This new nuclear reactor concept requires the development of computational tools to be applied in design and safety assessments to confirm improved inherent and passive safety features of this design. One approach to this issue is to modify the current computational codes developed for the simulation of Light Water Reactors towards their applicability for the new designs. This paper reports on the performed modifications of the TRACE system code to make it applicable to LFR safety assessments. The capabilities of the modified code are demonstrated on series of benchmark exercises performed versus other safety analysis codes. (Author)
[en] Importance measures are widely used for multiple purposes concerning safety improvement of nuclear power plants (NPPs). This paper proposes a new approach for performing the safety-related equipment prioritization for reliability centered maintenance purposes that considers two measures of risk importance of components based on a simplified Core Damage Frequency model derived from the plant specific level 1 PSA. It is also included the case of application to the Residual Heat Removal System (RHRS) at the Confrentes NPP in Spain
[en] The paper focuses on the optimization of the test and maintenance intervals under the criteria of unavailability or cost including the effect of the aging of the components and models of imperfect maintenance. The results obtained in the case of application, which focuses on a system of safety of a nuclear power station, show differences, mainly in the outage when you consider the aging. (Author)
[en] One goal of Generation IV reactors is to increase safety from those of previous generations. Different research platforms have identified the need to improve the reliability of the simulation tools to ensure the ability of the plant to accommodate the design basis transients established in preliminary safety studies. The article describes the modeling recirculation pumps in advanced sodium cooled reactors using TRACE codes. Upon implementation of the results obtained in models analyzing different design basis transients versus simplified approximations used in the reference models are compared.
[en] In a setting of stop with discovery of kernel crash is of special interest the maximum temperature of fuel elements (PCT) pod. The difficulty to directly measure this temperature makes is look for the measurement of the temperature of exit of the nucleus (CET). This paper proposes to study the correlation between these basic parameters of measurement in a commercial plant, based on the results of the simulation of different cases through the Thermo-hydraulic TRACE code.
[en] This paper presents the optimization of the testing surveillance and maintenance (TS and M) plan for high-pressure injection system (HPIS) in a nuclear power plant. Using the Particle Swarm Optimization (PSO) is obtained a group of viable solutions, each one corresponding to a non-dominated solution, which can be implemented in the plant.