Investigation of Effects of Piping Configuration and Water Supply Pressure on Air Intrusion

Author(s):  
Jaehyok Lim ◽  
Basar Ozar ◽  
Christopher E. Henry ◽  
Kevin B. Ramsden

An evaluation of the effects of geometry and water supply pressure on the void transport has been performed using RELAP5/MOD3.3 (patch03). Two different piping configurations were considered for a hypothetical nuclear power plant. The cases that were analyzed considered switchover between two different water supplies, i.e. condensate storage tank (CST) and essential service water system (SX) for a safety system that acted as the ultimate heat sink. In addition, two different pressures were considered for the pressure of SX to investigate the effect of supply water pressure on void transport. Results were interpreted based on the differences in the geometries of the piping configurations and supply water pressures.

Author(s):  
Shengtao Zhang ◽  
Ke Yi

Abstract Essential Service Water System (WES) is part of the nuclear power plant cooling system which provides the final heat sink for nuclear power plants. Therefore, WES must operate stably, safely and reliably for a long time. The total loss of WES accident is a design extended condition and will result in the loss of the final heat sink of the unit. The consequences of the accident are severe. In order to deal with the accident quickly and effectively and ensure the safety and economics of the power plant in accident condition, it’s necessary to formulate corresponding treatment strategy to deal with the transient. This paper developed a strategy for dealing with the total loss of WES with Residual Heat Removal System (RHR) not connected condition in Generation III nuclear power plant. The structure of the WES system and the types of failures that may occur are analyzed, and thus the symptoms of the faults are obtained and the entry conditions for the operating strategy are determined. The effect of faults on unit equipment and safety functions and the impact on nuclear steam supply system (NSSS) control are analyzed in this paper. Combined with the unit design, the system and equipment for controlling and mitigating related safety functions are analyzed, and the mitigation method and the fallback strategy of the fault are determined. Thereby a complete operating strategy of total loss of WES with RHR not connected is obtained. In addition, this paper analyzes and evaluates the operating strategy by simulating thermal hydraulic calculation for the first time. The results show that without staff intervention Component Cooling System (WCC) temperature reached 55°C limits after running a few minutes. Based on the intervention of the operating strategy proposed in this paper, WCC temperature reached the 55°C limits when the unit was operated at about 4 hours and 55 minutes. The result shows that and the strategy can effectively alleviate the failure and provide sufficient intervention time for the operator to bring the unit to a safe state.


Author(s):  
Li Nan ◽  
Lao Yi ◽  
Che Yinhui

When inspecting in the nuclear power plant, the bolt of the 001/004 pump in Essential Service Water system was found fracture. The bolt in 001 pump had ever fractured before, and it had been replaced. In this paper, the material, microstructure, energy dispersive spectrometry and mechanical check calculation of the bolt are analyzed. The result shows, the bolt breakage is for stress corrosion cracking, the corrosion element is Cl−. When the martensitic stainless steel is in the heat treatment, the temperature is improper control, which causing the Cr element distribution changed. So the ability of material to resist corrosion becomes poor which is the root cause of the bolt fracture.


Author(s):  
Yang Ting ◽  
Li Guang Sheng ◽  
Li Zeng Fen ◽  
Peng Yue ◽  
Hu Jian

For nuclear power stations, the main function of Essential Service Water System (ESWS) is to discharge the waste heat from reactor core and spent fuel pool to the environment controllably, which is directly related to the safety and economy of nuclear power stations. Usually ESWS use open water from sea, rivers, lakes, reservoirs, as heat transfer medium. Extremely harsh environmental conditions may disable system functions and even lead to ESWS failure, directly reduce the safety and economy of nuclear power stations, and cause serious nuclear accidents. Failure of ESWS is one of the main reasons that lead to the Fukushima nuclear accident because of the loss of electricity after the earthquake and tsunami. Based on the typical ESWS configuration and conditions of serving nuclear power stations in China, the influence of environmental conditions on the function of water system is studied, and the corresponding measures are analyzed. These conditions can be divided into three categories: temperatures, water levels, and physical and chemical characteristics. Temperatures affect cooling characteristic of ESWS mainly. Nuclear power stations in tropical areas need to focus on cooling capacity might be reduced by high temperature. Those in cold region need attention to excessive cooling and freezing problems caused by low temperature. The influence of water levels is mainly fluid transport capacity and selection of equipment to ESWS. When the range of natural water level is too wide, designers shall consider measures to narrow it, such as the construction of highly reliable reservoir. Inland nuclear power stations shall try to ensure the reliability of ESWS; prevent water level changes beyond the scope of design caused by drought and flood disasters. The effects of physical and chemical properties are derived from the open water characteristics, including high salinity, high chloride ion concentration, carrying solid particles, suspended solids, and aquatics, and so on. These characteristics will cause the equipment and pipeline eroded or even damaged, aqueducts of intake and output jammed, heat exchangers of the final heat sink weakened and other negative effects, resulting in ESWS performance decline. Some of these factors are the characteristics of station site natural environment, some others are changes caused by human activities. Some factors are sustained, long-term; some others may be sudden, temporary. Influence on these factors need to be taken measures from many aspects, including structure, biological disinfection, special materials and equipment, environmental protection measures around the nuclear power station, and so on. On the whole, the environmental factors that affect ESWS in the nuclear power stations are wide, and the influence mechanism is more complex. These factors ultimately act on ESWS, but most of them cannot be banished inside of ESWS or the final heat sink system. Against the negative effects from environmental conditions, it has to be considered from all steps in the engineering of nuclear power stations, including design, construction and operation. All the measures shall be suitable to local conditions, in order to ensure the safety and economy of nuclear power stations.


Author(s):  
Longze Li ◽  
Mingjun Wang ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu

The severe accident of CPR1000 caused by station blackout with the SG safety valve failure is simulated and analyzed using MELCOR code in this work. The CPR1000 power plant severe accident response process and the results with three different assumptions, which are no the seal leakage nor the auxiliary feed water, the seal leakage and auxiliary feed water exist, the seal leakage exist but no auxiliary feed water separately, are analyzed. According to the calculation results, without the seal leakage and auxiliary feed water, pressure vessel would fail at 9576 s. When auxiliary feed water was supplied, pressure vessel’s failure time would delay nearly 30000s. When the seal leakage exists, pressure vessel’s failure time would delay about 50 s. The results are meaningful and significant for comprehending the detailed process of severe accident for CPR1000 nuclear power plant, which is the basic standard for establishing the severe accident management guideline.


Author(s):  
Dominique Lagrange ◽  
Vincent Venturini ◽  
Georges Bezdikian ◽  
Jacques Salaun

In Electricite´ de France (EDF) probabilistic analyses for Reactor Pressure Vessel (RPV) life management, it’s used to take the temperature of the Reactor Spent fuel Pit cooling and treatment System storage tank as a constant. The aim of our study is to evaluate the stability of this temperature. Since 1999, we collected for several sites, and several nuclear plant units, the temperature of the Reactor Spent fuel Pit cooling and treatment System storage tank. Our results illustrate that this temperature depends on the season and the site. We first proposed to give a modelisation of this temperature dependent of the external temperature; even if this modelisation leads to a good R2, it’s not optimum. We also proposed to explain the temperature using the temperature of the essential service water or for lack the temperature of the river. In the case of extreme quantile study (meaning low temperature), we proposed to use the normal approximation, which seems to be conservative.


Author(s):  
Shinichi Matsuura ◽  
Ichiro Tamura

It is important in the confirmation of the safety of the nuclear power plant to clarify the response behavior of a vertical cylindrical water storage tank under seismic motion. When a vertical cylindrical tank is shaken by a large earthquake, deformation of side shell due to the elephant foot buckling, the oval vibration etc. may occur. The occurrence of those deformations depends on materials, shapes, stored water level and time history of seismic motion. Then, response behavior was obtained for a condensate storage tank (CST) model under large seismic motion such as standard earthquake Ss multiplied by 2 with the elastic-plastic finite element calculation. In this calculation, dynamic water pressure and elastic-plastic characteristics of the material were taken into account. In this case, the elephant foot bulge did not occur but the oval vibration of side shell became dominant. Based on the result, we estimated the structural integrity of the tank.


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