scholarly journals Research on the Ventilation System Valve Control and Power Distribution of a Nuclear Power Plant Conventional Island

2021 ◽  
Vol 2005 (1) ◽  
pp. 012156
Author(s):  
Jiaye Shao ◽  
Li Li ◽  
Ze Song
2018 ◽  
Vol 57 (4) ◽  
pp. 3483-3491
Author(s):  
H.A. Refaey ◽  
A.M. Refaey ◽  
M.M. Kandil ◽  
M.A. Moawed

Author(s):  
Jianfeng Yang ◽  
Handing Wang ◽  
Xiaoming Zhang ◽  
Bingchen Feng ◽  
Weijin Wang ◽  
...  

According to the research of the operating principle, installation position and running environment of the 380VAC emergency electrical power distribution cabinets (Hereinafter referred to as electrical cabinets) of a nuclear power plant in China, there are three aspects caused by earthquake that seriously affect the safety of the electrical cabinets, including relay chatter, failure of electrical cabinet structure and spatial interactions. Relay chatter refers to contacts of the relay being changed during the period of strong shaking. It may lead to associated circuits malfunction and the equipment failure of the relay control unless it can be effectively reset. The purpose of relay chatter is to find out these relays whose consequences are unacceptable after earthquake and calculate failure probability. Failure of electrical cabinet structure in the earthquake is to carry out seismic fragility evaluation. The goal of seismic fragility evaluation is to assess a given value which describes the ground acceleration capacity and the corresponding uncertainties, and then, the conditional probability of failure as a function of peak ground acceleration [PGA] and a family of fragility curves can be obtained. In this paper, finite element model of the electrical cabinet is established using ANSYS Workbench software. According to the electric cabinets seismic failure mode, we take some of the parameters including the parameters of the floor response spectrum, material strength parameters and so on as the input to calculate the median ground acceleration capacity and the corresponding uncertain parameters. The seismic spatial interactions are defined as the electrical cabinet destroyed due to the surrounding objects failure by falling, collapse, etc. Therefore, if necessary, it is needed to evaluate the seismic fragility of the surrounding objects. Usually through walking down, checking the design drawings or the combination of the above methods, we can find out the surrounding objects for an electric cabinet. So we analyze the seismic risk of the electrical cabinet from the above three aspects. When the results of the above three aspects obtained, we convolute of the electrical cabinet fragility with the seismic hazard curve which represents the frequency of occurrence of earthquake motions at various levels of intensity at the site. Then Monte Carlo sampling is adopted to analyze the uncertainty distribution. In this article, Risk Spectrum Professional software (reference 8) and Risk Spectrum Hazard lite software (reference 9) are used to complete the calculation and get some quantitative seismic risk insights. The above seismic risk insights can support the establishment of seismic probabilistic safety analysis model (Hereinafter referred to as SPSA) for a nuclear power plant, which helps to formulate seismic improvement strategies.


2016 ◽  
Vol 31 (3) ◽  
pp. 207-217
Author(s):  
Ghonche Baghban ◽  
Mohsen Shayesteh ◽  
Majid Bahonar ◽  
Reza Sayareh

An accurate analysis of the flow transient is very important in safety evaluation of a nuclear power plant. In this study, analysis of a WWER-1000 reactor is investigated. In order to perform this analysis, a model is developed to simulate the coupled kinetics and thermal-hydraulics of the reactor with a simple and accurate numerical algorithm. For thermal-hydraulic calculations, the four-equation drift-flux model is applied. Based on a multi-channel approach, core is divided into some regions. Each region has different characteristics as represented in a single fuel pin with its associated coolant channel. To obtain the core power distribution, point kinetic equations with different feedback effects are utilized. The appropriate initial and boundary conditions are considered and two situations of decreasing the coolant flow rate in a protected and unprotected core are analyzed. In addition to analysis of normal operation condition, a full range of thermal-hydraulic parameters is obtained for transients too. Finally, the data obtained from the model are compared with the calculations conducted using RELAP5/MOD3 code and Bushehr nuclear power plant data. It is shown that the model can provide accurate predictions for both steady-state and transient conditions.


2018 ◽  
Vol 72 ◽  
pp. 01001
Author(s):  
Qiang Li ◽  
Jia-qing Zhang ◽  
Jin-mei Li ◽  
Shi-jing Ren

This paper takes the nuclear power plant power distribution cabinet fire as example and adopts FPRA methodology to analyse the failure probability of cable insulation at contiguous region among distribution cabinet. The influential factor that fire brigade dealt with the fire in the spot is considered in this research. The cumulative distribution function of fire brigade’s arriving and extinguishing time is conducted discretization, confirming the sample statistics of areas and establishing the numbers of cable insulation failure events of distribution cabinet fire. Combining failure predicted model of cable insulation and calculating the failure situation of cable insulation under each of fire background, obtaining the probability that cable insulation failure events among the distribution cabinet fires. The results show that the HRRmax of distribution cabinet fire in term of case scenario is 750 kW, the cable never happens the insulation failure; when the distribution cabinet fire is 1000 kW, the possibility of cable insulation failure is 12.6%.


2022 ◽  
Vol 2022 ◽  
pp. 1-13
Author(s):  
Izza Shahid ◽  
Nadeem Shaukat ◽  
Amjad Ali ◽  
Meer Bacha ◽  
Ammar Ahmad ◽  
...  

A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.


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