Nuclear Power Plants In China Research Articles

A model of 90Sr distribution in the sea near Daya Bay Nuclear Power Plant in China

The impact on the environment of radionuclide release from nuclear power plants has attracted increased attention, especially after the accident at Fukushima Daiichi Nuclear Power Plant in Japan. Based on the mechanisms of adsorption/desorption at solid/liquid interfaces and a surface micromorphology model of sediments, a theoretical expression of the distribution coefficient K d is derived. This coefficient has significant effects on the distribution of radionuclide in seawater, suspended sediment and seabed sediment. K d is then used to simulate 90Sr transport in the sea near the Daya Bay Nuclear Power Plant. The simulation results are compared with field measurements of tidal level, current velocity, suspended sediment concentration and 90Sr concentrations in the same period. Overall, the simulated results agree well with the field measured data. Thus, the derived expression for K d is capable of interpreting realistic adsorption/desorption processes. What’s more, conclusion is drawn that about 40% 90Sr released by Daya Bay Nuclear Power Plant will be adsorbed by suspended sediment and 20% by seabed sediment, only about 40% 90Sr will remain in the sea near Daya Bay Nuclear Power Plant in South China Sea.

A design of human–machine interface for the two-modular high-temperature gas-cooled reactor nuclear power plant

A design of human–machine interface (HMI) has been developed to ensure the safety and availability of the two-modular high-temperature gas-cooled reactor nuclear power plant in China. As the first design for the plant with two modules coupled to one steam turbine, the staffing arrangement and HMIs are different from current nuclear power plants (NPPs). One control room is used to monitor and control the two reactor modules, the turbine generator, and the balance-of-plant (BOP). Considering the digital control room style, the elements of the HMIs in the control room are the large display panels and control consoles. According to the function analysis, the HMIs can be divided into three continuous parts: HMIs of reactor #1, conventional island, and reactor #2, from left to right for three operators respectively, and HMIs for the two reactor modules are basically the same. A Distributed control system (DCS) is employed for the non-safety controls, and computerized HMIs on the consoles are the interfaces for frequent use. The necessary safety operations and displays are set on the safety console of each reactor, independently. The large display panels are designed as an overview of the whole plant, which consist of large screen displays, mimic displays and alarm tiles. The failure modes and the diversities of HMIs are analyzed. Some special human factor engineering (HFE) strategies are considered for the control authorities of the two modules, the control of the systems shared between the two modules, etc. Two different solutions for the arrangement of HMIs and staffing are discussed. For lack of operating experiences, a verification platform is built to make verification and validation (V&V) of the HMIs. After sufficient validation, the design will be realized in the demonstration construction of the high-temperature gas-cooled reactor-pebble bed module (HTR-PM) in China.

Failure analysis on abnormal wall thinning of heat-transfer titanium tubes of condensers in nuclear power plant Part I: Corrosion and wear

Titanium tubes used in condensers in a nuclear power plant in China encountered abnormal wall thinning, and was thus forced to temporarily stop operation or it could bring about catastrophic safety problems. Most of the wall thinning happened at quite regular positions on the tubes and these failure tubes were located similarly in the condensers, indicating some common problems. To find out the root cause and mechanism of the thinning failure, we conducted surface deposit analysis, appearance inspection, microstructure analysis and composition analysis of the samples by means of X-ray diffraction (XRD), stereo microscope, scanning electron microscope (SEM) and Energy Dispersive Spectrometer (EDS). The results revealed that the wall thinning was primarily caused by eccentric contact wear and three-body contact wear rooted in processing defect of internal borings, corrosion products deposit and sagging, and foreign particles. Finally, countermeasures were proposed for repair and prevention.

Radioactivity of Drinking-Water in the Vicinity of Nuclear Power Plants in China Based on a Large-Scale Monitoring Study

The public concern for radioactivity of drinking-water has been increasing in recent years after the rapid development of nuclear power plants, and especially after the Fukushima nuclear accident. In this study, the radioactivity of water samples collected in the vicinity of nuclear facilities from seven provinces in China was measured and an average annual equivalent effective dose derived from drinking-water ingestion was calculated. The results showed that, in winter and spring, the activities of gross α and β ranged from 0.009 Bq/L to 0.200 Bq/L and from 0.067 Bq/L to 0.320 Bq/L, respectively. While, in summer and autumn, the activities of gross α and β varied from 0.002 Bq/L to 0.175 Bq/L and from 0.060 Bq/L to 0.334 Bq/L. Our results indicated that the gross α and β activities in these measured water samples were below the WHO recommended values (0.5 Bq/L for gross α and 1.0 Bq/L for gross β) and the annual equivalent effective dose derived from drinking-water ingestion was at a safe level.

Effect of the Fukushima nuclear accident on the risk perception of residents near a nuclear power plant in China

We assessed the influence of the Fukushima nuclear accident (FNA) on the Chinese public's attitude and acceptance of nuclear power plants in China. Two surveys (before and after the FNA) were administered to separate subsamples of residents near the Tianwan nuclear power plant in Lianyungang, China. A structural equation model was constructed to describe the public acceptance of nuclear power and four risk perception factors: knowledge, perceived risk, benefit, and trust. Regression analysis was conducted to estimate the relationship between acceptance of nuclear power and the risk perception factors while controlling for demographic variables. Meanwhile, we assessed the median public acceptable frequencies for three levels of nuclear events. The FNA had a significant impact on risk perception of the Chinese public, especially on the factor of perceived risk, which increased from limited risk to great risk. Public acceptance of nuclear power decreased significantly after the FNA. The most sensitive groups include females, those not in public service, those with lower income, and those living close to the Tianwan nuclear power plant. Fifty percent of the survey respondents considered it acceptable to have a nuclear anomaly no more than once in 50 y. For nuclear incidents and serious incidents, the frequencies are once in 100 y and 150 y, respectively. The change in risk perception and acceptance may be attributed to the FNA. Decreased acceptance of nuclear power after the FNA among the Chinese public creates additional obstacles to further development of nuclear power in China and require effective communication strategies.

A New Application of Support Vector Machine Method: Condition Monitoring and Analysis of Reactor Coolant Pump

Fukushima nuclear power plant accident caused huge losses and pollution and it showed that the reactor coolant pump is very important in a nuclear power plant. Therefore, to keep the safety and reliability, the condition of the coolant pump needs to be online condition monitored and fault analyzed. In this paper, condition monitoring and analysis based on support vector machine (SVM) is proposed. This method is just to aim at the small sample studies such as reactor coolant pump. Both experiment data and field data are analyzed. In order to eliminate the noise and useless frequency, these data are disposed through a multi-band FIR filter. After that, a fault feature selection method based on principal component analysis is proposed. The related variable quantity is changed into unrelated variable quantity, and the dimension is descended. Then the SVM method is used to separate different fault characteristics. Firstly, this method is used as a two-kind classifier to separate each two different running conditions. Then the SVM is used as a multiple classifier to separate all of the different condition types. The SVM could separate these conditions successfully. After that, software based on SVM was designed for reactor coolant pump condition analysis. This software is installed on the reactor plant control system of Qinshan nuclear power plant in China. It could monitor the online data and find the pump mechanical fault automatically.

Discussion on IAEA and China Safety Regulation for NPP Coastal Defense Infrastructures against Typhoon/Hurricane Attacks

The World Meteorological Organization estimates that about 90 percent of all natural disasters is extreme meteorological hazards like typhoon/hurricane and tropical cyclone triggered disasters. With the increasing tendency of natural hazards, the typhoon induced surge, wave, precipitation, flood and wind as extreme external loads menacing Nuclear Power Plants (NPP) in coastal and inland provinces of China. For all of the planned, designed and constructed NPP in China the National Nuclear Safety Administration of China and IAEA recommended Probable Maximum Hurricane/ Typhoon/(PMH/T), Probable Maximum Storm Surge (PMSS), Probable Maximum Flood (PMF), Design Basis Flood (DBF) as safety regulations recommended for NPP defense infrastructures. This paper discusses the joint probability analysis of simultaneous occurrence typhoon induced extreme external hazards and compared with IAEA 2003-2011 recommended safety regulations for some NPP along China coast to make safety assessment based on the “As Low As Reasonable Practice” (ALARP) principle.

Failure analysis of leakage on titanium tubes within heat exchangers in a nuclear power plant. Part II: Mechanical degradation

AbstractSerious failure incidents like clogging, quick thinning, and leakage frequently occurred on lots of titanium tubes of heat exchangers in a nuclear power plant in China. In the Part I of the whole failure analysis study with totally two parts, factors mainly involving three kinds of electrochemical corrosions were investigated, including galvanic corrosion, crevice corrosion, and hydrogen‐assisted corrosion. In the current Part II, through microscopically analyzing the ruptures on the leaked tubes by scanning electron microscopy (SEM) and energy dispersive spectrometry (EDS), another four causes dominantly lying in the aspect of mechanical degradation were determined – clogging, erosion, mechanical damaging, and fretting. Among them, the erosion effect was the primary one, thus the stresses it exerted on the tube wall were also supplementarily evaluated by finite element method (FEM). Based on the analysis results, the different degradation extents and morphologies by erosion on the tubes when they were clogged by different substances such as seashell, rubber debris, and sediments were compared, and relevant mechanisms were discussed. Finally, countermeasures were put forward as well.

Managing construction risks of AP1000 nuclear power plants in China

Large and complex construction projects face risk from various sources and the successful completion of such projects depends on effective risk management. This study investigates the risk faced by Chinese firms participating in constructing AP1000 nuclear power plants in China. AP1000 nuclear reactors are new, Generation III+ reactors designed by Westinghouse and to be built first in China. The semi-structured interview approach is used to elicit information from experts involved in the AP1000 projects in China. Based on the interviews, various sources of risk are identified. In addition to general risks that megaprojects normally face, there are unique risks that arise from various sources such as technological, political, organizational, and individual personnel risks. Risk management strategies are proposed to manage general and unique risks identified in the study. The findings of this study would be helpful for Chinese companies involved in the construction of AP1000 nuclear power plants to mitigate the risks associated with the projects.

Blade Stress of the Reactor Coolant Pump of 300 MWe Nuclear Power Plant in China Based on Fluid-solid Coupling Method

In order to compute and analyze the blade stress in reactor coolant pump of a 300 MWe nuclear power plant in China,from the viewpoint of safety performance,the fluid-solid coupling technique is adopted to solve the fluid and solid coupling equation.Theoretical analysis shows that the blade stresses mainly include tensile stress caused by centrifugal force,bending and torsional stresses induced by flow-field pressure,and thermal stress brought by temperature field.From the analysis results,it is concluded that the max equivalent(Von-Mises) stress exists in fixed supports and the blade stress does not have strict periodicity,and the disordered distributions of blade stress lies in the complex internal flow.The differences between impeller and guide vane are proved by comparison of theory and simulation and the main factors are explained.Lastly the intensity examination proves that the intensity of the reactor coolant pump meets the requirement of ASME.This could be used for improving aerofoil design and ensuring performance and intensity.

Development of a thermal-hydraulic analysis software for the Chinese advanced pressurized water reactor

A point reactor neutron kinetics model, a drift-flow U-tube steam generator model, a non-equilibrium three-region pressurizer model and other models were established and a transient analysis code with Visual Fortran 6.5 has been developed to analyze the thermal-hydraulic characteristics of the Chinese advanced pressurized water reactor (AC-600). Visual input, real-time processing and dynamic visualization output were achieved with Microsoft Visual Studio.NET 2003, which greatly facilitate applications in the engineering. The software were applied to analyze the transient thermal-hydraulic characteristics of the loss of feed-water accident, the double loops loss-of-flow accident, the reactivity insertion accident, the sudden increase of feed-water temperature accident and the loss of offsite power accident for the Qinshan nuclear power plant in China. The obtained analysis results are significant to the improvement of design and safety operation of the plant.

Perception of people for the risk of Tianwan nuclear power plant

A questionnaire survey of residents’ risk perceptions related to Taiwan nuclear power plant in China was carried out to explore the determining factors that affect individual risk perception. This study proposed to pursue a more comprehensive understanding of factors that affected individual risk perception to nuclear power plants. Covariance structure analysis was conducted using risk perceptions of nuclear power as dependent variable and including interest and knowledge levels of nuclear power, acceptability, benefit perception, trust in nuclear power operation, and trust in government as independent variables. The use of the hypothesis of Elaboration likelihood model (ELM) was also proposed. The results showed that persons with higher levels of interest and knowledge of nuclear power had their own perceptions of risk closely associated with acceptability and potential benefits of nuclear power. In contrast, persons with no interest in and knowledge of nuclear power would have risk perceptions related to their trust in nuclear operation and the government, which partially supported the ELM hypothesis. All these results indicated that the government in China plays an important role in rational risk perceptions, and well-designed communication of risks will help the public to be involved in risk management and improve people’s rational acceptance of risk.

Geomagnetically Induced Currents in the High-Voltage Power Grid in China

With the operation of long transmission lines, unknown severe vibration and noise in transformers have sometimes been observed at the Shanghe substation and at the Ling'ao nuclear power plant in China. This paper gives a comparison of magnetic-storm recordings and measured transformer neutral current data and confirms that the disturbances were caused by geomagnetically induced currents (GICs). On November 7 and 9, 2004, the maximum GIC at the Ling'ao nuclear power plant was 47.2 A and 75.5 A, respectively, which are both larger than the neutral DC current caused by the HVDC monopole operation. The analysis of GIC data indicates that the GIC level is mainly determined by the geomagnetic-field variation rate, but the earth conductivity and power grid parameters are also important factors affecting GIC values. The data show that the GIC level is higher at the Ling'ao nuclear power plant than at the Shanghe substation. The reason is considered to be related to the grid structure and to the coast effect. Furthermore, because of smaller resistances, longer transmission lines, a larger scale, and the adoption of single-phase transformers, the influences of GIC on the ultra-high voltage power grid being constructed in China need to be studied more urgently.

Accident consequence assessment and its implications for emergency planning of the pilot commercial HTR plant in China

An effort has been launched in China to construct a pilot commercial plant using high temperature reactors. This paper presents the major results of accident consequence assessment performed in the framework of a preliminary environmental impact evaluation. Based on this, the implications of this consequence calculation for emergency planning are discussed. Finally, the current regulatory requirements on emergency zoning around nuclear power plants in China and some discussions about the establishment of emergency planning zones around the Taiwan Nuclear Power Plant imported from Russia are also briefly described in this paper.

Some lessons learned from the development and updating processes of the probabilistic safety assessments of Daya Bay nuclear power plant

The quality of probabilistic safety assessment (PSA) should be commensurate with the application for which it is intended and the role the PSA results play in the integrated decision making process. The updating of the PSA model would be ineffective if it is not initiated according to the requirements of the applications. This paper summarizes the lessons learned from the PSA development and updating processes in Daya Bay Nuclear Power Plant (NPP) in China. While some of the issues are specific, the discussion of these quality issues will be helpful to the NPPs in China and maybe in other countries.

UPTAKE OF 137Cs AND 90Sr IN RICE PLANTS

In order to study the transfer of water-borne 137Cs and 90Sr into rice plants, pot experiments were done by growing rice plants on soils taken from rice beds near two nuclear power plants: Qinshan Nuclear Power Plant and Daya Bay Nuclear Power Plant in China. The vertical migration of 137Cs and 90Sr in the soil samples was studied, and results show that almost all the contaminants were retained in the top 1 cm of soil with at most a few percent down to 3 cm of soil. The areal transfer factors of 137Cs and 90Sr were studied by adding various activity concentrations of the two radionuclides to the irrigation water and then measuring the activity concentration in the various rice tissues. The areal transfer factors were also analyzed for their dependence on soil characteristics, time and duration of the contamination. The areal transfer factors obtained in this study refer to the absorption of the fallout nuclides deposited in the irrigation water in a rice field as opposed to other studies that deal with uniform activity concentration in the soils.

Uptake of137Cs and90Sr in rice grown on soils from Qinshan and Daya Bay area

In order to study the transfer of water borne 137 Cs and 90 Sr into rice plants, pot experiments were done by growing rice plants on soils taken from rice beds near two nuclear power plants : Qinshan Nuclear Power Plant and Daya Bay Nuclear Power Plant in China. The vertical migration of 137 Cs and 90 Sr in the soil samples due to leaching was studied, and results show that almost all the contaminants were retained in the top 1 cm of soil, with at most a few percent down to 3 cm of soil. The transport factors (TF) of 137 Cs and 90 Sr were studied by adding various activity concentrations of the two radionuclides to the irrigation water and then measuring the activity concentration in the various rice tissues. The TF were also analysed for their dependence on soil characteristics, time and duration of the contamination.

Flow structure of the plane turbulent impinging jet in cross flow

This paper presents both experiment and numerical simulation on Two-dimensional vertical impinging jet in cross flow. Intake pond of the cooling water system is a requisite component of thermal or nuclear power plant. But in intake pond sediment deposition is heavy. Impinging jet was equipped in a nuclear power plant in China and has a significant impact on sediment suspension. However, before this investigation, little attention has been paid to the mechanism of this application case. In response to the demands of the engineering design, started this study. Experiments are performed for different Reynolds number ratio of the jet-to-cross-flow R varying from 0.13 to 0.22. The measurement results of turbulent parameter and flow structure are obtained. Moreover the turbulent parameters play an important role on sediment transportation. As well the experiment results agree well with the calculations provided from κ - ε turbulence model.

Adaptation of COSYMA and assessment of accident consequences for Daya Bay nuclear power plant in China

The program package COSYMA for assessing the radiological and economic consequences of nuclear accidents, developed with the support of the European Commission, was applied to investigate the health effects and risks from accidental releases of radioactive material from the Daya Bay nuclear power plant. Population distribution data in the range of 80 km around the site and hourly meteorological data for the year 1985 representative of accident consequence analysis were used. The results showed that early effects are more important at distances closer to the site, while the number of fatal cancers is closely related to the population density and the late effects are still important at distances larger than 50 km from the site. The mean annual expected values for early mortality and late mortality estimated for the population within a circle of 80 km around the Daya Bay nuclear power plant are 4.5×10 −3 and 0.1 yr −1, respectively.

HIs classification in HCR model for operator reliability study of nuclear power plants

The human cognition reliability model is used to study time-related human response interaction (HI). The classification of HI types is an important base. At present, the two types of HIS classification (i.e. SRK classification and CP classification) are based on subjective or objective characteristics respectively by which some phenomena can be hardly explained. Since the types were limited, they could not meet the requirements. The integrate approach of HIs classification is compatible with the original methods, and HIs are classified according to subjective and objective characteristics. The integrate approach contains more detailed types and has higher precision. It is more advantageous and convenient for the classification of HIs. The integrate approach has been used in the study of operator reliability for nuclear power plants in China. The results were satisfactory, and by which the integrate approach was verified.