Abstract
A loss-of-coolant accident of nuclear power plant is simulated using the system analysis code. The analytical condition is based on the international standard problem 26 conducted with the integral effect test facility. The experimental analysis is also performed, and the result of plant analysis is compared with the experiment and the experimental analysis. The discharge coefficient of critical flow model is determined so as to obtain the agreement of reactor pressure between the experiment and the experimental analysis, and is used for the plant analysis. The thermal-hydraulic phenomena in the experiment such as core heat up are simulated well by the two analyses, while some problems in the experiment for simulating plant accidents are made clear. Parametric sensitivity analyses of plant accidents are performed as an example of reliable safety evaluation using the validated plant model.
Highlights
Loss of coolant accidents (LOCAs) are one of the most significant accidents among the design basis accidents of nuclear reactors, and many research works have been performed so far including experiments and numerical analyses
The results of plant system analyses are used as the boundary conditions for the detailed thermal-hydraulic analyses based on the computational fluid dynamics approach [5] and for the transient structural analyses [6]
Several code users including engineers of software companies are generally involved in these analyses, since the vast knowledge of plant and integral effect tests (IETs) systems such as the design, operating conditions and control procedures is necessary for developing analytical models
Summary
Loss of coolant accidents (LOCAs) are one of the most significant accidents among the design basis accidents of nuclear reactors, and many research works have been performed so far including experiments and numerical analyses. Several code users including engineers of software companies are generally involved in these analyses, since the vast knowledge of plant and IET systems such as the design, operating conditions and control procedures is necessary for developing analytical models. The accident data were limited to several plant parameters, but many measurement data were obtained by the experiment Both the plant accident analysis and the experimental analysis were performed using the TRAC code [12], and the thermal-hydraulic phenomena during the plant accident were shown to be simulated well by the LSTF experiment. The analysis of power plant cold-leg break LOCA with the condition of ISP 26 is performed. Analytical Model The PWR plant analysis and the LSTF experimental analysis are performed using the RELAP code [1]. The Henry-Fauske critical flow model [1] is applied with the discharge coefficient of 0.75 according to the preliminary calculations
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