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Verification of Mitigation Capability for Total Loss of Feedwater Accident and Sensitivity Study on Feed-and-Bleed Cooling in a CPR1000 Nuclear Power Plant 认领 引用
1
作者 Bo Zhang Zhenhua Zhang 《Energy Engineering》 EI 2026年第9期530-547,共18页
Total loss of feedwater accident is a typical transient among the design extension conditions for pressurized water reactor nuclear power plants,directly related to the loss of core cooling capability.Chinese nuclear ... Total loss of feedwater accident is a typical transient among the design extension conditions for pressurized water reactor nuclear power plants,directly related to the loss of core cooling capability.Chinese nuclear safety regulations require in-depth analysis of such conditions,while most of the Generation II and modified Generation II units currently in operation were designed prior to the issuance of these regulatory requirements,and their mitigation capability remains to be verified.To evaluate the mitigation capability of CPR1000 nuclear power units in operation in China for this accident,an accident sequence involving main feedwater pump trip combined with complete failure of the auxiliary feedwater system under full power condition was simulated based on a high-fidelity simulation platform.By strictly following the emergency operating procedures for operator interventions,the transient responses of key safety parameters including primary coolant system pressure,coolant inventory,core outlet temperature,fuel temperature,and containment pressure were analyzed.The results show that,under the synergistic effect of automatic system actions and procedure guidance,the operator manually opened all three sets of pressurizer safety valves at about 60 min,establishing the feed-and-bleed cooling mode.The minimum reactor pressure vessel water level was 67.7%,the core remained uncovered,and both the peak fuel cladding temperature and core outlet temperature remained below the limits.At about 98 min,the residual heat removal system entry conditions were satisfied,and the core was successfully brought to a safe state.A sensitivity study on the influence of the number of opened safety valve sets was performed.The results indicate that when only two sets of safety valves were successfully opened,all acceptance criteria were still satisfied;when only one set was opened,the heat removal capacity of the safety valves was insufficient,with the pressure exhibiting saw-tooth fluctuations in the range of 2.3-3.7 MPa(a),and the residual heat removal system entry conditions could not be satisfied,thus making a smooth transition to a safe state difficult.These results provide an engineering reference for similar units in coping with total loss of feedwater accidents and for optimizing the feed-and-bleed cooling operational strategy. 展开更多
关键词 Nuclear power plant total loss of feedwater mitigation capability verification feed-and-bleed cooling sensitivity study
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Lifespan Prediction of Electronic Card in Nuclear Power Plant Based on Few Samples 认领 引用
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作者 XU Yong CAI Yunze SONG Lin 《Journal of Shanghai Jiaotong university(Science)》 EI 2025年第6期1188-1194,共7页
A lifespan prediction model was developed based on a few samples to provide decision-making information for operation and maintenance,as well as improve the economy and safety of nuclear power plant(NPP)operations.Thi... A lifespan prediction model was developed based on a few samples to provide decision-making information for operation and maintenance,as well as improve the economy and safety of nuclear power plant(NPP)operations.This paper applies a Weibull model to forecast the lifespan of electronic cards with a few samples in NPPs.Relationship between the lifespan prediction of electronic cards and the ambient temperature is revealed using the Arrhenius equation.Censored samples are used to compensate for the lack of fault electronic card data.Scale parameter and shape parameter of the Weibull model are optimized by adjusting the weight ratio between the censored data and the fault data.Characteristic life is then obtained using the rank regression fitting equation.Parameters of the Arrhenius equation can be calculated by dividing the samples into groups according to the ambient temperature.A case study of the intermediate range high-voltage electric card of ex-core neutron detectors demonstrates that the lifespan prediction of electronic cards in NPPs can be successfully predicted with a few samples by combining the Weibull model and the Arrhenius model.This can help provide preventive maintenance recommendations for electronic cards.Finally,operation suggestions for the electronic card’s ambient temperature can be made by utilizing the temperature-life model. 展开更多
关键词 lifespan few samples Weibull model Arrhenius equation nuclear power plant(NPP)
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Observation and Assessment of Heat Transfer Coefficient of Thermal Discharge for Coastal Nuclear Power Plants 认领 引用
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作者 ZHU Qichao HUANG Chuanjiang +4 位作者 XIA Yuzhuo YANG Ying GUO Jingsong XIA Changshui QIAO Fangli 《Journal of Ocean University of China》 SCIE CAS CSCD 2025年第2期281-288,共8页
The heat transfer coefficient of the water surface is an important parameter in the design of thermal discharge in nuclear power plant engineering.In this study,in situ observations were performed in the northwestern ... The heat transfer coefficient of the water surface is an important parameter in the design of thermal discharge in nuclear power plant engineering.In this study,in situ observations were performed in the northwestern South China Sea near a coastal nuclear power plant to evaluate the applicability of heat transfer coefficient calculation algorithms commonly used in marine thermal discharge engineering in China.The results show that the Regulation for Hydraulic and Thermal Model in Cooling Water Projects(SL 160-2012)is not applicable in calculating the heat transfer coefficient in offshore areas.SL 160-2012 significantly overestimates the heat loss at the sea surface.However,Code for Design of Cooling for Industrial Recirculating Water(GB/T 50102-2014)performs well,and its estimation coefficient is roughly consistent with the estimations of the COARE 3.6 bulk algorithm,which is extensively used in physical oceanography for calculating air-sea heat fluxes,and the Gunneberg formula.In a 3-day observation,the average heat transfer coefficients estimated using these three algorithms were 50.4,48.5,and 48.8 W m-2℃-1,respectively,with a deviation of less than 4% among them,whereas that estimated using SL 160-2012 was as high as 176.3 W m-2℃-1.The abnormally large value of SL 160-2012 is due to its additional cooling term,which is artificially increased by 100 times because of the incorrect unit conversion used when developing the regulation.If this error is corrected,the value will decrease to 50.5 W m-2℃-1,which is very close to the estimation of GB/T 50102-2014. 展开更多
关键词 heat transfer coefficient thermal discharge sea-air heat flux temperature rise nuclear power plant COARE algorithm
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Using the response surface method to conduct wave hazard assessment for a floating nuclear power plant 认领 引用
4
作者 Shu-Wen Yu Xin-Yan Xu Chang-Hong Peng 《Nuclear Science and Techniques》 SCIE EI CAS CSCD 2025年第6期187-200,共14页
A floating nuclear power plant(FNPP)is an offshore facility that integrates proven light-water reactor technologies with floating platform characteristics.However,frequent contact with marine environments may lead to ... A floating nuclear power plant(FNPP)is an offshore facility that integrates proven light-water reactor technologies with floating platform characteristics.However,frequent contact with marine environments may lead to wave-induced vibrations and oscillations.This study aimed to evaluate the wave danger on FNPPs,which can negatively impact FNPP functionality.We developed a hydrodynamic model of an FNPP using potential flow theory and computed the frequency-domain fluid dynamic responses.After verifying the hydrodynamic model,we developed a predictive model for FNPP responses.This model utilizes a genetic aggregation methodology for batch prediction while ensuring accuracy.We analyzed all the wave data from a selected sea area over the past 50 years using the constructed surrogate model,enabling us to identify dangerous marine areas.By utilizing the extreme value distribution of important wave heights in these areas,we determined the wave return period,which poses a threat to FNPPs.This provides an important method for analyzing wave hazards to FNPPs. 展开更多
关键词 Floating nuclear power plant Wave hazard Hydrodynamic model
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Numerical Simulation of Hydrodynamic Forces Acting on Curved Shape Trash Intercepting Net in Nuclear Power Plant Under Regular Waves 认领 引用 被引量：4
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作者 WANG Bin WANG Xin-long +4 位作者 GAO Shan QIAO Dong-sheng LI Yu DENG Ziran OU Jin-ping 《China Ocean Engineering》 SCIE EI CSCD 2024年第6期983-998,共16页
The occurrence of blockages of trash intercepting net in nuclear power plant due to marine biofouling has become increasingly frequent,leading to significant changes in the mechanical state.This paper establishes a CF... The occurrence of blockages of trash intercepting net in nuclear power plant due to marine biofouling has become increasingly frequent,leading to significant changes in the mechanical state.This paper establishes a CFD(Computational Fluid Dynamics)model to simulate the hydrodynamic forces of trash intercepting net under the action of regular waves.The porous media model is used to calculate the hydrodynamic forces,and the maximum mooring load is also evaluated.The simplified calculation method considering the different curved shape based on the flat nets are proposed,and the influences of wave parameters,solidity,and curved shape are investigated.The results indicate that under the regular wave conditions,as the solidity increases,the phenomenon of secondary wave peaks becomes more pronounced.The horizontal wave force reduction coefficient follows a three-piecewise linear relationship with the non-dimensional deformation level of curved shape.The trash intercepting net exhibits more potent scattering effects on short-wave conditions,displaying significant non-linear characteristics.The deformation level of the trash intercepting net is a significant factor influencing the mooring load. 展开更多
关键词 trash intercepting net curved shape CFD model drag forces mooring load nuclear power plant
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Investigation on Flow Accelerated Corrosion Mitigation for Secondary Circuit Piping of the Third Qinshan Nuclear Power Plant 认领 引用 被引量：5
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作者 ZHAO Liang HU Jianqun +1 位作者 WU Zhigang WANG Kin 《Chinese Journal of Mechanical Engineering》 SCIE EI CAS CSCD 2011年第2期214-219,共6页
Flow accelerated corrosion(FAC)is the main failure cause of the secondary circuit carbon steel piping in nuclear power plants.The piping failures caused by FAC have resulted in numerous unplanned outages and tragic fa... Flow accelerated corrosion(FAC)is the main failure cause of the secondary circuit carbon steel piping in nuclear power plants.The piping failures caused by FAC have resulted in numerous unplanned outages and tragic fatalities.The existing researches focus on the main factors contributing to FAC,which include metallurgical factors,environmental factors and hydrodynamic factors.Some effective FAC management methods and programs with long term monitoring and inspection data analysis are recommended.But a comprehensive FAC management system should be developed in order to mitigate and manage FAC systematically.In this paper,the FAC influencing factors are analyzed in combination with the operating conditions of the secondary circuit piping in the Third Qinshan Nuclear Power Plant(TQNPP),China(Third Qinshan Nuclear Power Company Limited,China).A comprehensive FAC mitigation and management system is developed for TQNPP secondary circuit piping.The system is composed of five processes,viz.materials substitution,water chemical optimization,long-term monitor strategy for the susceptible piping,integrity evaluation of the local thinning defects,and repair or replacement.With the implementation of the five processes,the material of FAC sensitive pipe fittings are modified from carbon steel to stainless steel,N_2H_4 and NH_3 are finally selected as the water chemical regulator of secondary circuit,the secondary circuit pips are classified according to FAC susceptibility in order to conduct long term monitoring strategy,and an integrity evaluation flow for local thinning caused by FAC in carbon steel piping is developed.If the component with local thinning defects is not fit-for-service,corresponding repair or replacement should be conducted.The comprehensive FAC mitigation and management system with five interrelated processes would be a cost-effective method of increasing personnel safety,plant safety and availability. 展开更多
关键词 flow accelerated corrosion nuclear power plant secondary circuit piping
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Research on the attribution evaluating methods of dynamic effects of various parameter uncertainties on the in-structure floor response spectra of nuclear power plant 认领 引用 被引量：3
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作者 Li Jianbo Lin Gao +1 位作者 Liu Jun Li Zhiyuan 《Earthquake Engineering and Engineering Vibration》 SCIE EI CSCD 2017年第1期47-54,共8页
： Consideration of the dynamic effects of the site and structural parameter uncertainty is required by the standards for nuclear power plants （NPPs） in most countries. The anti-seismic standards provide two basic m... ： Consideration of the dynamic effects of the site and structural parameter uncertainty is required by the standards for nuclear power plants （NPPs） in most countries. The anti-seismic standards provide two basic methods to analyze parameter uncertainty. Directly manually dealing with the calculated floor response spectra （FRS） values of deterministic approaches is the first method. The second method is to perform probability statistical analysis of the FRS results on the basis of the Monte Carlo method. The two methods can only reflect the overall effects of the uncertain parameters, and the results cannot be screened for a certain parameter＇s influence and contribution. In this study, based on the dynamic analyses of the floor response spectra of NPPs, a comprehensive index of the assessed impact for various uncertain parameters is presented and recommended, including the correlation coefficient, the regression slope coefficient and Tornado swing. To compensate for the lack of guidance in the NPP seismic standards, the proposed method can effectively be used to evaluate the contributions of various parameters from the aspects ＆sensitivity, acuity and statistical swing correlations. Finally, examples are provided to verify the set of indicators from systematic and intuitive perspectives, such as the uncertainty of the impact of the structure parameters and the contribution to the FRS of NPPs. The index is sensitive to different types of parameters, which provides a new technique for evaluating the anti-seismic parameters required for NPPs. 展开更多
关键词 uncertain parameter floor response spectra （FRS） soil-structure interaction （SSI） seismic analysis andstructural design nuclear power plant （NPP）
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Probabilistic Approach of Coastal Defense Against Typhoon Attacks for Nuclear Power Plant 认领 引用 被引量：1
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作者 刘德辅 韩凤亭 +3 位作者 庞亮 李华军 谢波涛 王风清 《China Ocean Engineering》 SCIE EI 2010年第2期265-275,共11页
With the global warming and sea level rising, it is widely recognized that there is an increasing tendency of typhoon occurrence frequency and intensity. The defenses code against typhoon attacks for nuclear power pla... With the global warming and sea level rising, it is widely recognized that there is an increasing tendency of typhoon occurrence frequency and intensity. The defenses code against typhoon attacks for nuclear power plant should be calibrated because of the increasing threat of typhoon disaster and severe consequences. This paper discusses the probabilistic approach of definitions about ＂probable maximum typhoon＂ and ＂probable maximum storm surge＂ in nuclear safety regulations of China and has made some design code calibrations by use of a newly proposed Double Layer Nested Mtdti-objective Probability Model （DLNMPM）. 展开更多
关键词 typhoon nuclear power plant coastal engineering code calibration double layer nested multi-objective probability model
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Development and Application of Maintenance Template in Pressurized Water Reactor Nuclear Power Plant 认领 引用 被引量：2
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作者 张圣 陈宇 +1 位作者 曹智鹏 莫春铌 《Journal of Donghua University(English Edition)》 EI CAS 2015年第1期162-165,共4页
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
关键词 pressurized water reactor(PWR) nuclear power plant maintenance template maintenance program maintenance optimization
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Evaluation of protective quality of prestressed concrete containment buildings of nuclear power plants 认领 引用 被引量：1
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作者 郑砚国 李惠强 《Journal of Central South University》 SCIE EI CAS 2011年第1期238-243,共6页
The permeability and sorptivity properties of the two prestressed concrete containment buildings （PCCBs） of a nuclear power plant in South China, which had been under operation for 5 years, were measured by using th... The permeability and sorptivity properties of the two prestressed concrete containment buildings （PCCBs） of a nuclear power plant in South China, which had been under operation for 5 years, were measured by using the autoclam permeability system. The air permeability, sorptivity and water permeability indexes of No.1 PCCB are smaller than or equal to 0.11 ln（102 Pa）/min, 0.98×10 ^7 m3/minl/2 and 1.93×10 ^7 m3/min1/2, respectively, and the air permeability, sorptivity and water permeability indexes of No.2 PCCB are smaller than or equal to 0.17 In（102 Pa）/min, 1.6×10 ^7 m3/min1/2 and 4.43 ×10 ^7 m3/min1/2, respectively. Based on the criteria for evaluating the protective quality of concrete structures in terms of their permeability and sorptivity properties, proposed by the research organization of the autoclam permeability system, the protective quality of No. 1 PCCB is still in very good grade and that of No.2 PCCB is not in very good grade but in good grade, and the in-service inspection of the protective quality of No.2 PCCB should be strengthened in the future. 展开更多
关键词 nuclear power plant prestressed concrete containment building permeability properties sorptivity properties
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Effects of Liaoning Hongyanhe Nuclear Power Plant on the Zooplankton Community in Summer of 2017 认领 引用
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作者 WANG Junjian TAO Zhencheng +3 位作者 WANG Yantao WEI Hao LIU Xin LI Chaolun 《Journal of Ocean University of China》 SCIE CAS CSCD 2020年第5期1140-1152,共13页
To evaluate the effects of the Hongyanhe Nuclear Power Plant on the zooplankton community in the surrounding seawater during summer, multiple environmental factors and zooplankton distribution along the east coast of ... To evaluate the effects of the Hongyanhe Nuclear Power Plant on the zooplankton community in the surrounding seawater during summer, multiple environmental factors and zooplankton distribution along the east coast of Liaodong Bay were investigated in the summer of 2017. In particular, the influences of seawater temperature, salinity, and chlorophyll a(Chl a) on the zooplankton community were analyzed. Zooplankton abundances and Chl a concentrations along the east coast of Liaodong Bay showed an initial increase followed by a decrease from July to September. During the three months, the zooplankton abundance was the highest(8116.70 ind m-3) in August. The Shannon-Wiener index showed a downtrend from July to September, with the average value falling from 1.65 in July to 1.50 in September. Calanus sinicus, Paracalanus parvus, copepodid, and bivalve larvae were the dominant species/groups in the three months. The effects of the nuclear power plant's outlet on the environment factors were mainly reflected in the increased seawater temperature. Redundancy analysis showed that the zooplankton community was jointly affected by seawater temperature, salinity and Chl a concentration, and the degree of this impact varied monthly. The impact of seawater temperature on the zooplankton community was stronger than that of salinity. The primary impact of the Hongyanhe Nuclear Power Plant on the structure and distribution of the zooplankton community in the surrounding seawater during the summer was increased seawater temperature, which caused a reduction in the abundance of dominant species/groups. 展开更多
关键词 zooplankton abundance community structure diversity environmental factor Hongyanhe Nuclear Power Plant
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Image Quality Improvement for Underwater Visual Inspections of Nuclear Power Plants 认领 引用
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作者 HUANG San’ao WANG Xudong +1 位作者 LIANG Ying XU Ke 《Transactions of Nanjing University of Aeronautics and Astronautics》 EI CSCD 2021年第6期1037-1047,共11页
Visual inspection of the key components of nuclear power plants(NPPs)is important for NPP operation and maintenance. However,the underwater environment and existing radiation will lead to image degradation,thus making... Visual inspection of the key components of nuclear power plants(NPPs)is important for NPP operation and maintenance. However,the underwater environment and existing radiation will lead to image degradation,thus making it difficult to identify surface defects. In this study,a method for improving the quality of underwater images is proposed.By analyzing the degradation characteristics of underwater detection image,the image enhancement technology is used to improve the color richness of the image,and then the improved dark channel prior(DCP)algorithm is used to restore it. By modifying the estimation formula of transmittance and background light,the correction of insufficient brightness in DCP restored image is realized. The proposed method is compared with other state-of-the-art methods. The results show that the proposed method can achieve higher scores and improve the image quality by correcting the color and restoring local details,thus effectively enhancing the reliability of visual inspection of NPPs. 展开更多
关键词 image quality improvement visual inspection nuclear power plant underwater image
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Development and Research of Non⁃Stirring Conveying Device for Waste Resin in Nuclear Power Plant 认领 引用
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作者 Jianfa Li Yongzhen Hua +2 位作者 Mingmei Liu Rui Zhang Taishan Lou 《Journal of Harbin Institute of Technology(New Series)》 CAS 2023年第3期45-59,共15页
Blockage in the storage and transportation of waste resin is a difficult problem in the radioactive waste treatment process of nuclear power plants.In this study,in order to solve the problems of unstable resin transp... Blockage in the storage and transportation of waste resin is a difficult problem in the radioactive waste treatment process of nuclear power plants.In this study,in order to solve the problems of unstable resin transport concentration and easy blockage of conveying equipment and pipelines in nuclear power plants in China,a set of non⁃stirring conveying devices is developed,and theoretical calculations,simulation analysis and experimental verification are carried out.By transporting resin using the no stirring conveying device developed in this paper,it is not only to eliminate the risk of blockage and ensure the safety of transportation,but also to adjust the concentration of conveying resin to change the transport efficiency according to the operating conditions.The effective bearing rate of waste resin storage tank can be improved,so that the comprehensive performance of waste resin storage and transportation in nuclear power plants can be greatly improved. 展开更多
关键词 nuclear power plant ion exchange resin transportation no stirring device no blockage
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Nuclear power plant fault diagnosis based on genetic-RBF neural network 认领 引用 被引量：2
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作者 SHI Xiao-cheng XIE Chun-ling WANG Yuan-hui 《Journal of Marine Science and Application》 2006年第3期57-62,共6页
It is necessary to develop an automatic fault diagnosis system to avoid a possible nuclear disaster caused by an inaccurate fault diagnosis in the nuclear power plant by the operator. Because Radial Basis Function Neu... It is necessary to develop an automatic fault diagnosis system to avoid a possible nuclear disaster caused by an inaccurate fault diagnosis in the nuclear power plant by the operator. Because Radial Basis Function Neural Network (RBFNN) has the characteristics of optimal approximation and global approximation. The mixed coding of binary system and decimal system is introduced to the structure and parameters of RBFNN, which is trained in course of the genetic optimization. Finally, a fault diagnosis system according to the frequent faults in condensation and feed water system of nuclear power plant is set up. As a result, Genetic-RBF Neural Network (GRBFNN) makes the neural network smaller in size and higher in generalization ability. The diagnosis speed and accuracy are also improved. 展开更多
关键词 geneticalgorithm （GA） RBF neural network nuclear power plant
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A framework for nuclear power plant emergency response system 认领 引用
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作者 Wang-kun CHEN 《Nuclear Science and Techniques》 SCIE CAS 2010年第6期375-378,共4页
The purpose of this study is to establish an intelligent expert system for nuclear power plant emergency response.A new framework of environmental risk management methodology by the concept of pattern recognition was ... The purpose of this study is to establish an intelligent expert system for nuclear power plant emergency response.A new framework of environmental risk management methodology by the concept of pattern recognition was introduced in this paper.A knowledge-based decision support system for emergency response and risk management of nuclear power plant was also discussed.The mathematical pattern relationship of accidental release effects on neighboring area and the corresponding response measures were presented in this paper.With this decision system,the decision maker can specify the procedure and minimize their human error in the decision process.The improvement of risk response and the quality of management system could be upgraded by this system.Besides,the methodology can also be served as a basis for the future development of environmental risk response system design. 展开更多
关键词 Expert system Emergency response Risk management Nuclear power plant
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Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant 认领 引用 被引量：2
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作者 Yi Ping Wang Qingkang Kong Xianjing 《Earthquake Engineering and Engineering Vibration》 SCIE EI CSCD 2017年第1期55-67,共13页
Abstract： The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete... Abstract： The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel （PCCV） model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels. 展开更多
关键词 nuclear power plant prestressed concrete containment vessel aseismic safety analysis
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Decomposition of oil cleaning agents from nuclear power plants by supercritical water oxidation 认领 引用 被引量：1
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作者 Shi-Bin Li Xiao-Bin Xia +2 位作者 Qiang Qin Shuai Wang Hong-Jun Ma 《Nuclear Science and Techniques》 SCIE EI CAS CSCD 2022年第4期83-95,共13页
Oil cleaning agents generated from nuclear power plants(NPPs)are radioactive organic liquid wastes.To date,because there are no satisfactory industrial treatment measures,these wastes can only be stored for a long tim... Oil cleaning agents generated from nuclear power plants(NPPs)are radioactive organic liquid wastes.To date,because there are no satisfactory industrial treatment measures,these wastes can only be stored for a long time.In this work,the optimization for the supercritical water oxidation(SCWO)of the spent organic solvent was investigated.The main process parameters of DURSET(oil cleaning agent)SCWO,such as temperature,reaction time,and excess oxygen coefficient,were optimized using response surface methodology,and a quadratic polynomial model was obtained.The determination coefficient(R2)of the model is 0.9812,indicating that the model is reliable.The optimized process conditions were at 515 C,66 s,and an excess oxygen coefficient of 211%.Under these conditions,the chemical oxygen demand removal of organic matter could reach 99.5%.The temperature was found to be the main factor affecting the SCWO process.Ketones and benzene-based compounds may be the main intermediates in DURSET SCWO.This work provides basic data for the industrialization of the degradation of spent organic solvents from NPP using SCWO technology. 展开更多
关键词 Supercritical water oxidation Oil cleaning agent Nuclear power plants Response surface methodology
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PSA study of the effect of extreme snowfall on a floating nuclear power plant:case study in the Bohai Sea 认领 引用 被引量：1
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作者 Lan‑Xin Gong Qing‑Zhu Liang Chang‑Hong Peng 《Nuclear Science and Techniques》 SCIE EI CAS CSCD 2023年第11期212-225,共14页
This study presents a probabilistic safety analysis(PSA)method for the external event of extreme snowfall on a floating nuclear power plant(FNPP)deployed in the Bohai Sea.We utilized the Weibull and Gumbel extreme val... This study presents a probabilistic safety analysis(PSA)method for the external event of extreme snowfall on a floating nuclear power plant(FNPP)deployed in the Bohai Sea.We utilized the Weibull and Gumbel extreme value distributions to fit the collected meteorological data and obtained a hazard curve for the event of an extreme snowfall where the FNPP is located,providing a basis for the frequency of extreme snowfall-initiating events.Our analysis indicates that extreme snowfall primarily affects the ventilation openings of the equipment,leading to the failure of devices such as the diesel generators.Additionally,extreme snowfall can result in a loss of off-site power(LOOP).Therefore,the developed extreme snowfall PSA model is mainly based on the LOOP event tree,considering responses such as snowfall removal by personnel.Our calculations indicate a core damage frequency(CDF)of 1.13×10-10 owing to extreme snowfall,which is relatively low.The results of the cut-set analysis indicate that valve failures in the core makeup tank(CMT),passive residual heat removal system(PRS),and in-containment refueling water storage tank(IRWST)significantly contribute to the CDF. 展开更多
关键词 Floating nuclear power plant(FNPP) ACP100 Extreme snow PSA External hazard
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Study on evaluation system for Chinese nuclear power plants 认领 引用 被引量：2
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作者 李松柏 程建秀 《Journal of Harbin Institute of Technology(New Series)》 EI CAS 2006年第4期415-420,共6页
This paper analyzes the meaning, structure, function and assessment methods of a nuclear power plant evaluation system, and the similarities and differences among various assessment methods. Based on this research an ... This paper analyzes the meaning, structure, function and assessment methods of a nuclear power plant evaluation system, and the similarities and differences among various assessment methods. Based on this research an integrated and detailed suggestion is proposed on how to establish and improve internal and external evaluation systems for Chinese NPPs. It includes： to prepare and implement the nuclear power plant operational management program, to build an integrated performance indicator system, to improve the present audit system and conduct the comprehensive evaluation system, to set up and implement the integrated corrective action system, to position precisely the status of operation assessment of nuclear power plants, to conduct the assessment aetivilies on constructing NPP, to initiate the specific assessment in some important areas, to establish industrv performance indicator system, to improve the assessment methods, to share the assessment results, to select, cultivate and certify the reviewers, and to enhance interuational communication and cooperation. 展开更多
关键词 nuclear power plant evaluation system
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Research on Verification Method of Motor Startups in Nuclear Power Plants Based on Topology Recognition 认领 引用 被引量：1
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作者 Li Baozhu Dong Weijie Chen Chao 《系统仿真学报》 CAS CSCD 北大核心 2024年第12期2813-2823,共11页
There are many motors in operation or on standby in nuclear power plants,and the startup of group motors will have a great impact on the voltage of the emergency bus.At present,there is no special or inexpensive softw... There are many motors in operation or on standby in nuclear power plants,and the startup of group motors will have a great impact on the voltage of the emergency bus.At present,there is no special or inexpensive software to solve this problem,and the experience of engineers is not accurate enough.Therefore,this paper developed a method and system for the startup calculation of group motors in nuclear power plants and proposed an automatic generation method of circuit topology in nuclear power plants.Each component in the topology was given its unique number,and the component class could be constructed according to its type and upper and lower connections.The subordination and topology relationship of switches,buses,and motors could be quickly generated by the program according to the component class,and the simplified direct power flow algorithm was used to calculate the power flow for the startup of group motors according to the above relationship.Then,whether the bus voltage is in the safe range and whether the voltage exceeds the limit during the startup of the group motor could be judged.The practical example was used to verify the effectiveness of the method.Compared with other professional software,the method has high efficiency and low cost. 展开更多
关键词 power supply for nuclear power plant automatic topology recognition startup of group motor simplified direct power flow algorithm verification method
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