Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and econom...Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and economic operation for nuclear power plants( NPPs) especially for a large number of nuclear powers under construction. Based on the development and application of maintenance template in developed countries,and combining with reliability-centered maintenance( RCM) analysis results and maintenance experience data over the past ten years in domestic NPPs, the development process of maintenance template was presented for Chinese pressurized water reactor( PWR) NPP,and the application of maintenance template to maintenance program development and maintenance optimization combined with cases were demonstrated. A shortcut was provided for improving the efficiency of maintenance optimization in domestic PWR NPP,and help to realize a safe,reliable,and economic operation for domestic NPPs.展开更多
This paper presents a novel method to solve old problem of water level control system of pressurized water reactor (PWR) steam generator (SG) of nuclear power plant (NPP) .The level control system of SG plays an impo...This paper presents a novel method to solve old problem of water level control system of pressurized water reactor (PWR) steam generator (SG) of nuclear power plant (NPP) .The level control system of SG plays an important role which effects the reliablity,safty,cost of SG and its mathematical models have been solved.A model of the conventional controller is presented and the existing problems are discussed. A novel rule based realtime control technique is designed with a computerized water level control (CWLC) system for SG of PWR NPP.The performance of this is evaluated for full power reactor operating conditions by applying different transient conditions of SG′s data of Qinshan Nuclear Power Plant (QNPP).展开更多
Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied. The analyses were carried out with different water injection rates at different core damage ...Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied. The analyses were carried out with different water injection rates at different core damage stages. The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region. Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K, because the core is quenched and reflooded quickly. The water injection at the peak core temperature of 1900 K, the hydrogen generation rate increases at low injection rates of the water, as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate. At peak core temperature of 2100–2300 K, the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core. Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture. Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region. However, hydrogen is generated if water is injected into the molten pool, because steam serves to the crust supporting the molten pool. Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation. Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.展开更多
The transfer of pressurized water reactor(PWR)technology from France to China is an important event in the history of Sino-French scientific and technological relations.China has gradually achieved self-reliance in th...The transfer of pressurized water reactor(PWR)technology from France to China is an important event in the history of Sino-French scientific and technological relations.China has gradually achieved self-reliance in the field of PWR technology through the introduction and subsequent absorption of France's 900 MW reactors.Compared with the process of introducing and absorbing similar technology from the United States by France,China's experience has been more complicated.This circumstance reflects the differences in the nuclear power technology systems between the two countries.France's industrial strength and early acquisition of nuclear power technology laid a solid foundation for mastering PWR technology.On the other hand,although China established a weak foundation through the implementation of the"728 Project,"and tried hard to negotiate with France,the substantive content of the technology transfer was very limited.By way of the policy transition from"unhooking of technology and trade"to"integration of technology and trade,"China ultimately accomplished the absorption and innovation of PWR technology through the Ling'ao NPP.展开更多
Stress corrosion cracking(SCC) of an SA508-309 L/308 L–316 L dissimilar metal weld joint in primary pressurized water reactor environment was investigated by the interrupted slow strain rate tension tests following a...Stress corrosion cracking(SCC) of an SA508-309 L/308 L–316 L dissimilar metal weld joint in primary pressurized water reactor environment was investigated by the interrupted slow strain rate tension tests following a microstructure characterization. The 308 L weld metal shows a higher content of δ ferrite than the 309 L weld metal. In addition, no obvious Cr-depletion but carbides precipitation at δ phase boundaries was observed in both 308 L and 309 L weld metals. The slow strain rate tension tests showed that the SCC susceptibility of the base and weld metals of the dissimilar metal weld joint follows the order of SA508 < 308 L weld metal < the heat affected zone of 316 L base metal < 309 L weld metal.The higher SCC susceptibility of 309 L weld metal than that of 308 L weld metal is likely due to the lower content of δ ferrite. In addition, a preferential SCC initiation in the 309 L weld metal adjacent to 308 L weld metal is attributed to few carbides in this region.展开更多
A computer model has been developed for prediction of the pressure in the pressurizer under transient conditions. In the model three separate thermodynamic regions which are not required to be in thermal equilibrium h...A computer model has been developed for prediction of the pressure in the pressurizer under transient conditions. In the model three separate thermodynamic regions which are not required to be in thermal equilibrium have been considered. The mathematical model derived from the general conservation equations includes all of the important thermal-hydraulics phenomena occurring in the pressurizer, i.e., stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer, etc. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented model will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant’8 pressurizer performance.展开更多
文摘Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and economic operation for nuclear power plants( NPPs) especially for a large number of nuclear powers under construction. Based on the development and application of maintenance template in developed countries,and combining with reliability-centered maintenance( RCM) analysis results and maintenance experience data over the past ten years in domestic NPPs, the development process of maintenance template was presented for Chinese pressurized water reactor( PWR) NPP,and the application of maintenance template to maintenance program development and maintenance optimization combined with cases were demonstrated. A shortcut was provided for improving the efficiency of maintenance optimization in domestic PWR NPP,and help to realize a safe,reliable,and economic operation for domestic NPPs.
文摘This paper presents a novel method to solve old problem of water level control system of pressurized water reactor (PWR) steam generator (SG) of nuclear power plant (NPP) .The level control system of SG plays an important role which effects the reliablity,safty,cost of SG and its mathematical models have been solved.A model of the conventional controller is presented and the existing problems are discussed. A novel rule based realtime control technique is designed with a computerized water level control (CWLC) system for SG of PWR NPP.The performance of this is evaluated for full power reactor operating conditions by applying different transient conditions of SG′s data of Qinshan Nuclear Power Plant (QNPP).
基金Supported by National Basic Research Program of China (No.2009CB724301)
文摘Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied. The analyses were carried out with different water injection rates at different core damage stages. The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region. Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K, because the core is quenched and reflooded quickly. The water injection at the peak core temperature of 1900 K, the hydrogen generation rate increases at low injection rates of the water, as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate. At peak core temperature of 2100–2300 K, the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core. Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture. Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region. However, hydrogen is generated if water is injected into the molten pool, because steam serves to the crust supporting the molten pool. Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation. Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.
基金a phase study of a key project of the Fourteenth Five-Year Plan of the Institute for the History of Natural Sciences,Chinese Academy of Sciences:“A Comparative Study of the Sino-Foreign History of Scientific and Technological Innovation:The Road to Scientific and Technological Self-Reliance and Self-Improvement”,E2291J01。
文摘The transfer of pressurized water reactor(PWR)technology from France to China is an important event in the history of Sino-French scientific and technological relations.China has gradually achieved self-reliance in the field of PWR technology through the introduction and subsequent absorption of France's 900 MW reactors.Compared with the process of introducing and absorbing similar technology from the United States by France,China's experience has been more complicated.This circumstance reflects the differences in the nuclear power technology systems between the two countries.France's industrial strength and early acquisition of nuclear power technology laid a solid foundation for mastering PWR technology.On the other hand,although China established a weak foundation through the implementation of the"728 Project,"and tried hard to negotiate with France,the substantive content of the technology transfer was very limited.By way of the policy transition from"unhooking of technology and trade"to"integration of technology and trade,"China ultimately accomplished the absorption and innovation of PWR technology through the Ling'ao NPP.
基金supported by the National Natural Science Foundation of China(No.51571204)the support of the Youth Scientific and Innovation Research Team for Advanced Surface Functional Materials(Southwest Petroleum University,No.2018CXTD06)the Young Scholars Development Found of Southwest Petroleum University(No.201899010040).
文摘Stress corrosion cracking(SCC) of an SA508-309 L/308 L–316 L dissimilar metal weld joint in primary pressurized water reactor environment was investigated by the interrupted slow strain rate tension tests following a microstructure characterization. The 308 L weld metal shows a higher content of δ ferrite than the 309 L weld metal. In addition, no obvious Cr-depletion but carbides precipitation at δ phase boundaries was observed in both 308 L and 309 L weld metals. The slow strain rate tension tests showed that the SCC susceptibility of the base and weld metals of the dissimilar metal weld joint follows the order of SA508 < 308 L weld metal < the heat affected zone of 316 L base metal < 309 L weld metal.The higher SCC susceptibility of 309 L weld metal than that of 308 L weld metal is likely due to the lower content of δ ferrite. In addition, a preferential SCC initiation in the 309 L weld metal adjacent to 308 L weld metal is attributed to few carbides in this region.
基金Shanghai institute for Nuclear Engineering Research and Design
文摘A computer model has been developed for prediction of the pressure in the pressurizer under transient conditions. In the model three separate thermodynamic regions which are not required to be in thermal equilibrium have been considered. The mathematical model derived from the general conservation equations includes all of the important thermal-hydraulics phenomena occurring in the pressurizer, i.e., stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer, etc. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented model will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant’8 pressurizer performance.
