Steam Generator Modal Analysis of Nuclear Power Plant

Author(s):  
Lu Yan ◽  
Chu Qibao ◽  
Wang Qing ◽  
Fang Yonggang

A method for forming a simplified model of steam generator which will be used in reactor coolant loop analysis has been shown here, as well as the modal analysis to this simplified SG model. This modal analysis results and the results of the SG provided by NPP designer are compared together in order to prove the design correctness. The comparison shows that the two are basically consistent.

2005 ◽  
Vol 235 (23) ◽  
pp. 2477-2484 ◽  
Author(s):  
Seong Sik Hwang ◽  
Hong Pyo Kim ◽  
Joung Soo Kim ◽  
Kenneth E. Kasza ◽  
Jangyul Park ◽  
...  

2019 ◽  
pp. 119-126

Aplicación de la Teoría de Perturbación – Método Diferencial- al Análisis de Sensibilidad en Generadores de Vapor de Centrales Nucleares PWR-Caso Angra I Aplication of the Perturbation Theory- Differential Methodto Sensibility Análisis in PWR Nuclear Power Plant Steam Generator- Angra I Giol Sanders R, Andrade de Lima F, Marques A, Gallardo A, Bruna M, Zúñiga A Institución Peruano de Energía Nuclear Universidad Federal de Rio De Janeiro-Brasil DOI: https://doi.org/10.33017/RevECIPeru2011.0033/ RESUMEN En este trabajo basado en la tesis del Magíster Roberto Giol S. [1] presenta una aplicación del formalismo diferencial de la teoría de perturbación a un modelo termohidráulico homogéneo de simulación del comportamiento estacionario de uno de los generadores de vapor de la Central Nuclear tipo PWR Angra I del Brasil. Se desarrolla un programa de cálculo PERGEVAP tomando como base el código GEVAP de Souza[2]. El programa PERGEVAP permite realizar cálculos de sensibilidad de funcionales lineales (temperatura media del primario)y no lineales (flujo de calor medio a través de las paredes de los tubos del generador) con relación a las variaciones de ciertos parámetros termo-hidráulicos(flujo másico del primario, calor específico, etc), Los resultados obtenidos con este formalismo son luego comparados con los obtenidos del cálculo directo con el propio código GEVAP, pudiéndose verificar una excelente concordancia. Este método se muestra promisorio para efectuar cálculos repetitivos asociados al diseño y análisis de Seguridad de los componentes de las Centrales Nucleares. Descriptores: teoría de perturbación, método diferencial, sensibilidad, generador de vapor, central nuclear PWR. ABSTRACT This report presents an application of the differential approach of the perturbation theory to an homogeneous model of a PWR steam generator in the Angra 1 Nuclear Power Plan in Brazil under steady-state conditions. Program PERGEVAP was built fom the code GEVAP developed by Souza and allows sensitivity calculations of linear (average primary loop temperature) and non-linear (average heat flux) functionals due to variations in some thermo-hydraulics parameters (flow rate, specific heat, , etc). Results obtained with this approach are then compared with direct calculations performed using the GEVAP code, with excellent agreements. The method has good potential to treat repeated calculations needed in the design and safety analysis of the Nuclear Plant components. Keywords: perturbation theory, differential method, steam generator, PWR nuclear Power Plant.


2021 ◽  
Author(s):  
Yonglu Du ◽  
Haotian Li ◽  
Minrui Fei ◽  
Ling Wang ◽  
Pinggai Zhang ◽  
...  

2017 ◽  
Vol 2017 ◽  
pp. 1-13 ◽  
Author(s):  
Hao Shi ◽  
Qi Cai ◽  
Yuqing Chen

The best estimation process of AP1000 Nuclear Power Plant (NPP) requires proper selections of parameters and models so as to obtain the most accurate results compared with the actual design parameters. Therefore, it is necessary to identify and evaluate the influences of these parameters and modeling approaches quantitatively and qualitatively. Based on the best estimate thermal-hydraulic system code RELAP5/MOD3.2, sensitivity analysis has been performed on core partition methods, parameters, and model selections in AP1000 Nuclear Power Plant, like the core channel number, pressurizer node number, feedwater temperature, and so forth. The results show that core channel number, core channel node number, and the pressurizer node number have apparent influences on the coolant temperature variation and pressure drop through the reactor. The feedwater temperature is a sensitive factor to the Steam Generator (SG) outlet temperature and the Steam Generator outlet pressure. In addition, the cross-flow model nearly has no effects on the coolant temperature variation and pressure drop in the reactor, in both the steady state and the loss of power transient. Furthermore, some fittest parameters with which the most accurate results could be obtained have been put forward for the nuclear system simulation.


Author(s):  
Zhaohui Ren ◽  
Hui Ma ◽  
He Li ◽  
Guiqiu Song ◽  
Wenjian Zhou

The reactor coolant pump in nuclear power plant is the only revolving equipment in the nuclear power plant. Its functional stability will directly affect the security of nuclear power plant. The coolant pump of a very nuclear plant is examined by using response spectrum analysis to analysis dynamic characteristics and responses aiming at finding the natural frequencies of vibration, modes of vibration and seismic responses, and any possible step which may cause damage of the whole system. The favorable spectrum and unfavorable one are investigated as well. The paper focuses on avoiding the detrimental effects caused by earthquakes, therefore may lay down a theoretical foundation for structural design and installation.


2021 ◽  
Vol 152 ◽  
pp. 107945
Author(s):  
Jiuwu Hui ◽  
Jun Ling ◽  
He Dong ◽  
Gaixia Wang ◽  
Jingqi Yuan

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