Possibilities of generalization of experimental data on lifetime change of nuclear-power-plant parameters

Atomic Energy ◽  
1992 ◽  
Vol 73 (3) ◽  
pp. 762-764 ◽  
Author(s):  
V. S. Emel'yanov ◽  
O. G. Kamyshnikov ◽  
V. I. Morozkin ◽  
Yu. I. Raevskii
Keyword(s):  
Experimental Data ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power

scholarly journals The Use of System Codes in Scaling Studies: Relevant Techniques for Qualifying NPP Nodalizations for Particular Scenarios

2014 ◽  
Vol 2014 ◽  
pp. 1-13 ◽  
Author(s):  
V. Martinez-Quiroga ◽  
F. Reventos
Keyword(s):  
Experimental Data ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Analysis ◽  
Small Scale ◽  
Plant Operation ◽  
Essential Step ◽  
Hydraulic Safety ◽  
And Control

System codes along with necessary nodalizations are valuable tools for thermal hydraulic safety analysis. Qualifying both codes and nodalizations is an essential step prior to their use in any significant study involving code calculations. Since most existing experimental data come from tests performed on the small scale, any qualification process must therefore address scale considerations. This paper describes the methodology developed at the Technical University of Catalonia in order to contribute to the qualification of Nuclear Power Plant nodalizations by means of scale disquisitions. The techniques that are presented include the so-calledKv-scaled calculation approach as well as the use of “hybrid nodalizations” and “scaled-up nodalizations.” These methods have revealed themselves to be very helpful in producing the required qualification and in promoting further improvements in nodalization. The paper explains both the concepts and the general guidelines of the method, while an accompanying paper will complete the presentation of the methodology as well as showing the results of the analysis of scaling discrepancies that appeared during the posttest simulations of PKL-LSTF counterpart tests performed on the PKL-III and ROSA-2 OECD/NEA Projects. Both articles together produce the complete description of the methodology that has been developed in the framework of the use of NPP nodalizations in the support to plant operation and control.

TRACE Validation Against LOCA Transients Performed on FIX-II Facility

2013 ◽  
Author(s):  
Xiao Hu ◽  
Mian Xing ◽  
Weimin Ma
Keyword(s):  
Experimental Data ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Simulation Models ◽  
Primary System ◽  
Choked Flow ◽  
Loss Of Coolant Accident ◽  
Trace Model ◽  
Transient Thermal

As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. Loss-of-coolant accident (LOCA) is a kind of transient thermal hydraulic event which has been emphasized a lot as a most important threat to the safety of the nuclear power plant. The FIX-II experiments were performed to produce experimental data for understanding the initial stage of LOCA and so as to verify the computational codes. In the present study, based on FIX-II LOCA tests, simulation models for the tests of No. 3025, No. 3061 and No. 5052 which correspond to different LOCA cases were developed to validate the TRACE code (version 5.0 patch 2). The predictions of the TRACE code including the pressure in the primary system, the mass flow rate in certain key parts, and the temperature in the core were compared with the experimental data. The results show that TRACE model can well reproduce the transient thermal-hydraulic behaviors under different LOCA situations. In addition, sensitivity analysis are also performed to investigate the influence of particular models and parameters, including counter current flow limitation (CCFL) model and choked flow model on the results, which show that both the models have significant influence on the outcome of the model.

Life Extension of Bolted Flanged Connections From a Design Peculiarities Point of View

2004 ◽  
Author(s):  
Anatoly I. Efremov
Keyword(s):  
Experimental Data ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Point Of View ◽  
Life Extension ◽  
Design Improvement ◽  
Power Plant Equipment ◽  
Bolt Preload ◽  
Plant Equipment

Life extension of Bolted Flanged Connections (BFC) depends directly on early leakage that is a major cause of bolt degradation and failure. Experiments with standard BFC typically used in Nuclear Power Plant equipment and subjected to bolt preload and subsequent internal pressure revealed an influence of the BFC design peculiarities on early leakage. The experimental data became a basis for BFC design improvement.

Experimental Investigations on the Prototypic Steam Explosion

2005 ◽  
Author(s):  
J. H. Song ◽  
J. H. Kim ◽  
B. T. Min ◽  
S. W. Hong
Keyword(s):  
Experimental Data ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Test Section ◽  
Steam Explosion ◽  
Nuclear Power ◽  
Weight Percent ◽  
Experimental Investigations ◽  
Dynamic Force ◽  
Reference Case

This paper discusses results of a series of steam experiments using a prototypic material representing the molten core of nuclear power plant. Five experiments are discussed in detail in addition to a brief review of the previous experiments, where the focuses were on the effect of corium composition and external trigger on the strength of the steam explosion. A mixture of UO2:ZrO2 is used for the experiment, where the weight percent of each component is changed. One experiment was performed with corium at a composition of 70:30 without an external trigger as a reference case. Three tests were performed by using corium at the same composition with an external trigger. The last experiment was performed using corium at a composition of 80:20 with an external trigger. Various parameters are measured including the dynamic pressures on the wall of the test section and the dynamic force at the bottom of the test section. From the experimental data, the strength of the steam explosion was evaluated. It is shown that the strength of the steam explosion highly depends on the composition. A comparison between the cases with an external trigger and the cases without external trigger indicates that there is no substantial escalation of the strength of the explosion due to an external trigger.

Fuzzy multi-objective decision making approach for nuclear power plant installation

2020 ◽  
Vol 39 (5) ◽  
pp. 6339-6350
Author(s):  
Esra Çakır ◽  
Ziya Ulukan
Keyword(s):  
Linear Programming ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Energy Supply ◽  
Energy Demand ◽  
Nuclear Power ◽  
Power Plants ◽  
Total Duration ◽  
Objective Functions ◽  
Multi Objective

Due to the increase in energy demand, many countries suffer from energy poverty because of insufficient and expensive energy supply. Plans to use alternative power like nuclear power for electricity generation are being revived among developing countries. Decisions for installation of power plants need to be based on careful assessment of future energy supply and demand, economic and financial implications and requirements for technology transfer. Since the problem involves many vague parameters, a fuzzy model should be an appropriate approach for dealing with this problem. This study develops a Fuzzy Multi-Objective Linear Programming (FMOLP) model for solving the nuclear power plant installation problem in fuzzy environment. FMOLP approach is recommended for cases where the objective functions are imprecise and can only be stated within a certain threshold level. The proposed model attempts to minimize total duration time, total cost and maximize the total crash time of the installation project. By using FMOLP, the weighted additive technique can also be applied in order to transform the model into Fuzzy Multiple Weighted-Objective Linear Programming (FMWOLP) to control the objective values such that all decision makers target on each criterion can be met. The optimum solution with the achievement level for both of the models (FMOLP and FMWOLP) are compared with each other. FMWOLP results in better performance as the overall degree of satisfaction depends on the weight given to the objective functions. A numerical example demonstrates the feasibility of applying the proposed models to nuclear power plant installation problem.

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