bluuebunny/crossref_metadata_2025 huggingface dataset:
Feasibility Analysis and Demonstration of In-Vessel-Injection in the Early Stage of Severe Accident
Authors: Xia Shaoxiong , Zhan Dekui , Wu Zijie , Zhao Xinhai , Yu Yiqun
Published: November 2022

The core of the nuclear reactor will melt during severe accidents, which may damage the integrity of the reactor pressure vessel and containment, and release radioactive materials to the environment. The third-generation pressurized water reactor is equipped with severe accident mitigation systems to prevent the high temperature (3000K) corium from melting through the reactor pressure vessel. The severe accident mitigation system mainly includes: primary depressurization system, reactor pit flooding system, containment combustible system, containment heat removal system, etc.. Benefit from these systems, large release frequency (LRF) is restricted to a low level. However, these systems cannot stop the process of core degradation. If severe accident happens, the core will melt and the whole reactor cannot be reused again, causing irreversible economic losses. In order to improve the economy and safety of nuclear power plants, this paper proposes in-vessel-injection (IVI) in the early stage of severe accident, and proves the effectiveness of this measure. According to the analysis, in-vessel-injection in the early stage of severe accident can prevent the large-scale melting of the core during severe accident, and the risk of hydrogen and source items will be greatly reduced.
Research on Establishing Efficient Scheduling System of Spent Fuel From Chinese Nuclear Power Plants
Authors: Su Jiqiang , Liu Qun , Zhang Honglin , Shi Lei , Liu Hongjun , Li Yanrui , Hu Jian
Published: November 2022

As the number of nuclear power plants put into operation in China continues to increase, more and more spent fuel is produced. Before the spent fuel leaves the spent fuel pool to the final disposal of the spent fuel, it needs to go through multiple and various forms of transportation and storage. By establishing an efficient scheduling system for spent fuel, the efficient scheduling and overall coordination of each link is crucial for the safe management of spent fuel. This paper studies the connotation of the efficient scheduling system of spent fuel. Based on the enlightenment of the establishment of the efficient scheduling system for spent fuel in major nuclear power countries, it analyzes the necessity and current situation of establishing an efficient scheduling system for spent fuel in China’s nuclear power plants, and puts forward relevant measures and suggestions.
Digital Twin-Driven Development of Online Monitoring and Data Management Systems in TMSR-SF0
Authors: Liu Wen-Qian , Han Li-Feng , Huang Li
Published: November 2022

With the development of a new generation of information technology, the monitoring systems in the nuclear power field are also facing the technological revolution of digital reform. Based on the example of Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF0), this paper introduces the overall structure, modeling and mapping, and digital visualization expression means, then a complete development scheme of the real-time online 3D monitoring system is proposed. Firstly, the digital mapping model of the molten salt reactor was established, and the model docking and virtual scene rendering were completed in the Unity engine. Secondly, the back-end server using Experimental Physical and Industrial Control System (EPICS) and a proxy server using Node.js are established, and spatio-temporal data association is realized through Node-EPICS event driver and Socket.io. Finally, the MySql database interface and xCharts visualization framework are introduced to complete data storage and expression. Verified by practice, The 3D monitoring system is feasible in design and operation. Compared with the previous monitoring system, the data update period is increased to 20ms, and it has the whole process of data collection, network communication, graph element dynamic display, and other functions, which can promote the operation management and monitoring of nuclear power plants, and provide a reference for the digital transformation of monitoring technology in the nuclear power field.
Feasibility Study of Passive Water-Cooled Residual Heat Removal Scheme for Small Reactors
Authors: Yu Fei , Fu Yueming , Zhu Feng , Zhang Mingqian
Published: November 2022

With the development of nuclear energy and the need for environmental protection, some advanced small reactors have gradually entered the public’s field of vision. The passive natural circulation is an important part of the inherent safety feature for advanced small reactors. Based on the limitations of the site environment, the large heat exchange area of the air cooler, and the high cost of equipment construction for the air-cooled waste heat removal scheme of a small reactor, this paper studies the water-cooled passive residual heat removal scheme and analyzes the water-cooled water cooling by establishing a system natural circulation hydraulic calculation model. From the aspects of equipment scale, construction cost, and operation & maintenance cost, etc., the advantages and disadvantages of several schemes were analyzed.

This study shows that the water-cooled passive waste heat removal scheme can extract the core heat under accident conditions; it can reduce the equipment volume, and reduce the cost. Finally, suggestions are put forward for the follow-up improvement direction, and further research and realization of steam-water two-phase condensation heat transfer are made to improve heat exchange efficiency, reduce equipment scale, and improve the economy, which can provide a reference for the follow-up small reactor research and development.
Research on the Expected Carbon-14 Production and Discharge in Pressurized Water Reactors
Authors: Fu Pengtao
Published: November 2022

Carbon-14 is one of the most important radionuclides discharged to the environment from pressurized water reactors due to its long half-life and its important role in the biological chain. Carbon-14 is the largest contributor of the dose rate to the public from all radionuclides discharged to the environment during the normal operation of pressurized water reactors, and thus the production and discharge of Carbon-14 have been focused on in the industry in recent years.

Based on the generation mechanism in pressurized water reactor, one theoretical model of Carbon-14 generation has been established and the nitrogen concentration of all chemical species in the primary loops has been determined according to the measured ammonium. It predicts that the annual Carbon-14 generation in the primary loops is proportional to annual electricity output and the typical normalized Carbon-14 is 2.9E+02 GBq/GWe/yr. The theoretical model has been verified by the statistical analysis of annual Carbon-14 discharges from French PWR units. In addition, the quantity of Carbon-14 in radioactive solid waste has been estimated in these PWR units. It shows the generation of Carbon-14 in PWR cannot be effectively minimized because O-17 atoms, the predominant origin of Carbon-14, exist naturally in the primary loops during long-term operation. This approach can be applied to analyze the Carbon-14 production and discharges in operating pressurized water reactors and in the assessment of source term of the new pressurized water reactors.
The Characterization of Radiation Response for MAPS
Authors: Qin Zhiwei , Xu Shoulong , Dong Hanfeng
Published: November 2022

This paper studies the response of γ-ray ionizing radiation under the irradiation of single active Pixel Sensor in the irradiation chamber and the Calibration Chamber, Designs Radiation Experiments, and analyzes and discusses the characteristic values in different statistical regions of pixels in the bright and dark environment, the linear relationship and dispersion degree between the statistical eigenvalue and the dose rate under different statistical algorithms are studied. The experimental results show that in the dark environment, the background gray values of any region in the image are in the range of 10–30, and the eigenvalue have a good linear response. Under the bright environment, the global region of the image will be confused with the neighboring region because of the different background gray values of the color region in the Byer array, which will affect the characterization of the dose rate. In low dose rate, the signal of radiation response is too small, dark current and background dark noise produce interference signal, so the signal-to-noise ratio is too low, and 300 frames of image data are needed to achieve the detection accuracy. In high dose rate, the response signal is strong, and 2 consecutive frames of image data are needed to achieve the detection accuracy. The characteristic value of threshold pixel at low dose rate is good, and the characteristic value of threshold Pixel and total gray value at high dose rate are good, which meet the requirement of wide range detection. In the radiation field experiment, we can choose the corresponding algorithm to detect the dose rate according to the field environment. This study provides theoretical basis and data support for improving gamma-ray dose rate detection technology based on MAPS image and realizing radiation field dose rate characterization. In order to use the static environment to do the radiation experiment, we can design an algorithm to cut the edge of the frame image, and use the computer image processing technology to choose the area which is advantageous to the dose rate Characterization, To realize radiation Detection in dynamic environment.
Effects of Stoichiometry and High-Temperature Annealing on Zirconium Carbide Coating Layer in TRISO Particles
Authors: Cheng Xinyu , Liu Rongzheng , Liu Bing , Yang Xu , Liu Malin , Chang Jiaxing , Shao Youlin
Published: November 2022

Very-high-temperature gas-cooled reactors (VHTR) are being developed to provide higher thermal efficiency and high-temperature process heat. Zirconium carbide (ZrC) has been proposed as a potential coating material for TRistructural-ISOtropic (TRISO) coated fuel particles because of its excellent resistance to fission products corrosion, good thermal stability and higher mechanical strength under elevated temperatures. The integrity and performance of the ZrC coating of the TRISO particles are very important as it provides the main barrier for fission product release. Therefore, the microstructure and property evolution of ZrC coating deserve to be investigated. Fluidized-bed chemical vapor deposition (FB-CVD) has been conducted to fabricate the ZrC coating in a ZrCl4−C3H6-Ar-H2 system. The stoichiometry of ZrC was changed by controlling the feeding rate of ZrCl4 and the flow rate of C3H6. The ZrC coatings were annealed from 1700 °C to 2200 °C to study the possible changes in microstructures and temperature-dependent performances. The effect of stoichiometries on ZrC coating was studied by X-ray diffraction (XRD), scanning electron microscopy (SEM), Raman spectroscopy (Raman), and nanoindenter. Results showed that free carbon prevents grain growth under high-temperature annealing, and it reacts with ZrC1-x at higher temperatures to form pure phase ZrC. In addition, the microstructure evolution mechanism of ZrC at high temperatures was proposed.
Research and Treatment of Periodic Fluctuation of Steam Feed Pump Speed Regarding 900MW Nuclear Power Plant
Authors: Li Heng , Liu Xiaoyu , Wang Xufeng
Published: November 2022

A 900MW nuclear power unit has been in commercial use for 20 years. The steam-driven feed water pump uses a regulating system constructed by an analog board to control the speed of the pump. Since the analog board has been discontinued, and its reliability is reduced. The power plant has carried out digital transformation on the control part of the speed control system. Shortly after the renovation, the speed of the steam feed pump continued to fluctuate. The phenomenon accompanying the rotational speed fluctuation is a periodic fluctuation similar to a sine wave. The range of fluctuations varies from time to time, and is interspersed with non-sinusoidal fluctuations. When one of the feed pumps is placed in the manual control state, the speed fluctuation of the other pump will increase. After analysis, with the aging of the mechanical parts of the feed pump, the follow-up ability of the pump continues to weaken. Under the combined action of the four loops of speed, W/S differ., SG level and feed water flow, continuous fluctuations occur in the four parameters.

Aiming at the fluctuation problem, this paper relies on the real-time monitoring data of nuclear power plants and Matlab/Simulink to establish a nonlinear model and “Accompanying system” that reflect the actual working conditions on site, and conduct in-depth research on the mechanism of the problem by using control theory, system identification and “Accompanying system”, and propose solutions to resolve fluctuations. This method provides theoretical guidance for reducing or eliminating rotational speed fluctuations, thereby avoiding potential safety hazards caused by multi-parameter fluctuations in the operation of nuclear power units.
Study on Risk of Nuclear Spent Fuel Marine Transportation
Authors: Yiqing Liu , Wei Wang , Xiaodong Zhao , Zhong Gui , Jiangtao Chen
Published: November 2022

The safety of hazardous materials marine transportation is an important aspect of marine transportation safety. As the seventh category of hazardous materials, spent fuel has strong radioactivity and high decay heat, which makes the marine transportation of spent fuel concerned by the national competent department. Risk analysis is based on the identification of risk factors, through qualitative or quantitative analysis methods, to analyze the possible consequences of these risks and the possibility of these consequences, so as to determine the corresponding risk level and formulate targeted risk control measures. The safety of marine transportation depends not only on the safety of the cargo itself, but also on the navigation safety of the ship. On the basis of introducing the steps and methods of risk analysis, this paper selects various factors affecting the safety of spent fuel marine transportation from the aspects of cargo itself and ship navigation safety, and establishes a scientific and reasonable spent fuel marine transportation safety evaluation index system; The risk factors of maritime transportation are evaluated by risk evaluation methods such as expert scoring and risk coordinate diagram, some measures and suggestions are put forward according to the evaluation results.
Dynamic Modeling and Reactor Power Control of a Natural Circulation Pool-Type Lead-Cooled Fast Reactor
Authors: Feng Yun , Sun Pei-Wei , Wei Xinyu
Published: November 2022

The primary circuit of a natural circulation pool-type lead-cooled fast reactor adopts full natural circulation cooling technology, the reactor core power regulation process has the characteristics of long regulation time. When developing the reactor power control system, it is necessary to fully understand the dynamic characteristics of the reactor, so it is necessary to establish the nonlinear model of the natural circulation lead-cooled fast reactor. Based on the requirements of reactor dynamic characteristics research and power control system development, a nonlinear model of primary circuit is established on MATLAB/Simulink platform. The dynamic characteristics of the reactor are analyzed by introducing the reactivity disturbance of the reactor core and the flow and temperature disturbance of the secondary loop coolant. The transient response trend of the system is reasonable and conforms to the physical law, which can be used to analyze the dynamic characteristics of the system and the design of the control system. The core power control system adopts load following mode, and the load power measured value is used as the set value of the core thermal power for power regulation. When introducing the load power step change, the core thermal power change trend is consistent with the load power, and the control effect is satisfactory. The research results of this paper can provide reference for the design of power control system of natural circulation fast reactor.
Safety Analysis About Combined Heat and Power Generation of Nuclear Power Plant
Authors: Liu Shiyao
Published: November 2022

Nuclear energy is one of the most efficient, clean and stable energy sources. The technology of combined heat and power generation, which is widely used in the conventional thermal power plants all over the world, is also applicable to the nuclear power plants to make full use of the nuclear energy. But at present, the mode of combined heat and power generation is rarely used in the nuclear power plants in China, concerning about the nuclear safety problem. This study aims at the optimum use of nuclear energy to raise the potential margin of nuclear power plants, under the premise of ensuring nuclear safety, and analyzes the influence of the combined heat and power generation (such as the external supply of steam generated by nuclear power plants) of a pressurized water reactor on nuclear safety, including the influence on the normal operation, design basis conditions, design extension conditions and transient analysis. It shows that combined heat and power generation of nuclear power plants has no negative impact on nuclear safety. In the meanwhile, the mode of combined heat and power generation can improve the economy of nuclear power plants and has positive significance for the efficient use of nuclear energy.
First-Principles Study on Adsorption Reaction of Oxygen Molecules on Fe (110) Crystal Surface
Authors: Xiaochuan Zeng , Xuejun Li , Cuizhu He , Qiaodan Hu
Published: November 2022

The adsorption reaction between oxygen (O2) molecule and ferrum (Fe) (110) crystal surface in the oxidation process of Fe surface was studied by using the first-principles method. The differential charge density analysis of the adsorption sites of oxygen molecule on Fe (110) crystal surface, the calculation of adsorption energy at different sites and the analysis of electronic density of states showed that the stable adsorption position of oxygen molecule was parallel to Fe (110) crystal surface, and the oxygen atom tended to adsorb at the triangular gap of Fe atoms. The electronic structure of the adsorption system showed that the 2p electron orbital of oxygen atom plays a major role in the adsorption, and only O-Fe electron interaction exists when oxygen molecule is adsorbed in the parallel orientation, which makes the whole Fe (110) crystal surface lose electrons, increase the system potential and the risk of electrochemical corrosion. The research conclusions can provide theoretical support for the further insight in the oxidation corrosion mechanism of nuclear metal surface.
Numerical Investigation on Heat Transfer Features of Gas-Cooled Open Lattice Reactor in Normal Operations
Authors: Fang Junlin , Sun Qian , Ji Yu , Ye Zishen , Sun Jun , Sui Zhe
Published: November 2022

The gas-cooled reactor combined with Brayton cycle energy conversion system has been investigated with the application potential to meet the requirement of large power nuclear reactor systems in deep space explorations. Conceptual core design configured in the open lattice has attracted extensive attention due to the compact structure, and its heat transfer features need further investigation, including conduction, convection, and radiation. Many studies on thermal-hydraulic characteristics of similar structure reactors such as Pressurized Water Reactor (PWR) and Sodium Cooled Fast Reactor (SFR) have been carried out. However, different operating conditions bring the distinct heat transfer features. Referring to Prometheus program, this paper analyses the heat transfer features of gas-cooled open lattice reactor by numerical simulations. The computational fluid dynamic tool (ANSYS Fluent) is mainly used to investigate the effect of circumferential rod heat conduction, thermal radiation among fuel rods, and gas flow distribution. The results show that contributions of conduction, convection and radiation to the total reactor core heat transfer should be taken into account. The maximum temperature on the walls can be lower by increasing the rod thermal conductivity, enhancing the thermal radiation among fuel rods, and increasing the gas flow of inner core. This study will help to develop the system code for the thermal-hydraulic design and safety analysis of gas-cooled open lattice reactors.
Study on the Policy and Legislative System of Spent Nuclear Fuel Management in Different Developed Countries
Authors: Shi Lei , Chen Chao , Liu Hongjun , Su Jiqiang , Zhang Honglin , Liu Qun , Li Yanrui , Hu Jian
Published: November 2022

Spent nuclear fuel is an inevitable product from the development of nuclear energy. Almost all of the fuel content is radioactive, and long systematic process are required for the safety management, which has always been an important global issue. In order to make sure that spent nuclear fuel should be safely managed in different countries developing nuclear power, IAEA is establishing a sharing system of spent nuclear fuel management by concluding joint conventions and issuing safety standards. For different countries, the United States, France and Russia with nuclear power have all established a complete policy and legislative system for spent fuel management. In the US, policy decision of open-cycle has been made, and no commercial reprocessing is being conducted. In France and Russia, closed-cycle strategy is implemented with industrial-scale reprocessing plant in operation. At present, China has become the country with the largest scale of nuclear power under construction in the world. There will be a large number of spent nuclear fuel requiring properly and safely managed. The lessons-learning of how developed countries managing spent nuclear fuel arising is important for China. The authors suggest that it is necessary to combine the top-level design to the legal practice, so that there are laws to respect during all steps of spent fuel management, and responsibilities of all parties are clear.
Die Design and Process Study of Zr-4 Alloy Profile Extrusion
Authors: Qi Peng , Gao Bo , Zhang Zhihao , Yin Jianming
Published: November 2022

According to the application requirements of Zr-4 profile with outer square and inner circle for small nuclear reactors, the extrusion die structure and extrusion process parameters are optimized through numerical simulation and calculation. Based on the study of alloy friction boundary conditions and heat exchange conditions, the influences of die inlet angle and working belt structure size on metal flow behavior were clarified. Combined with the analysis of related factors such as flow stress and temperature rise variation law in extrusion process, determine the structure size of the die and the process parameter range of the extrusion, use the on-site 1250T horizontal extruder to carry out the pilot scale experiment of profile extrusion, verify the results of the preliminary analysis and calculation, determine the optimal process technology plan, and trial-produce The required outer square and inner circle zirconium alloy profiles that every sizes and performances meet the design requirements. The cross-sectional size of this profile is 36mm × 36mm, the inner hole diameter is 20mm, and the length is 1500∼2500(mm). It is the first zirconium alloy profile produced by the extrusion process in China. This process can ensure the grains are broken under large deformation, and obtain profiles with good performance, uniform wall thickness and longer length. At the same time, it can also reduce material loss, improve production efficiency.
Study on Forging Technology of Integrated Topcover Forging for AP1000 Reactor Pressure Vessel
Authors: Li Zhaoyin , Dong Kai , Tao Zhiyong
Published: November 2022

AP1000 topcover forging, as a key component in the Reactor Pressure Vessel (RPV), is located at the top of the RPV and is an important reactor primary component within the pressure boundary of the primary circuit, and its integrity is of great significance to the safe operation of nuclear power equipment. The traditional topcover part is welded by three parts: the upper head, the topcover flange and the stacking tube socket. When the components are in service, the in-service inspection workload of the equipment is large, and the safety risk is high under complex working conditions. Therefore, the study of a large forging integrated top cover with a stacking tube socket is carried out to realize the upper head, the topcover flange and the stacking tube as integrated topcover. The integral forging of the integrated topcover is an urgent problem to be solved.

This paper mainly studies the development of forging process for topcover forging of AP1000 RPV with SA508 III material. Based on the shape characteristics and forging rheological characteristics of the topcover forgings, the topcover tube seat was identified as the key point of forming, so the concave round cake preform was designed, and the final forging was formed by spinning the preform. The preform forming simulation and the final spinning forming simulation were carried out by DEFORM software respectively. The simulation results show that the metal filling is good, the forging allowance is sufficient, but the fracture factor of the bottom head is high. In order to more comprehensively verify the feasibility of the forming scheme, Trial production of 1:20 small scale simulation parts was carried out, and the forming results showed that the small scale parts of the integrated topcover were well filled after spinning. Through the manufacture of theoretical and practical simulation parts, the forging process scheme of spinning preforms through two sets of dies is verified, which effectively solves the problem that the size of the top cover forgings is difficult to guarantee. It was manufactured according to the designed forging process using the 300-ton series of steel ingots, and the final product size and non-destructive testing met the requirements, and the qualified product was delivered.
Common Cause Failure Analysis and Countermeasure of Digital Instrument and Control System in Nuclear Power Plant
Authors: Wang Ping , Hu Lisheng
Published: November 2022

The design philosophy of defence in depth for nuclear power plant establishes in consideration of technical, artificial and organizational management failures. Defence in depth principle is a safety design idea that runs through all stages of production activities in the whole life of nuclear power plants. Each of defence lines must be reliable and independent of each other as much as possible. According to the system function and operation requirements of nuclear power plants, this paper analyses latest requirements of defence in depth and diversity line of nuclear power plants, based on the research and analysis of laws and regulations, standard and good practices at domestic and abroad. According to the requirements of independence, common cause fault and diversity, the independence and diversity of defence line in depth of instrument and control systems are analyzed, the weak points of common cause fault of defence line in depth of instrument and control system are identified, and the improved analysis method and improved realization steps are put forward. And then each of its lines of defence provided diversity for common cause failure. This paper discusses the types of common cause faults and typical countermeasures of defence in depth and diversity line for digital instrument and control system. This paper summarizes the common cause failure analysis methods, processes and countermeasures for digital instrument and control systems in nuclear power plant.
The Numerical Method Research for Simulating the SMR Steady-State Primary System Flow and Head Loss
Authors: Ma Zhongying , Li Ge , Wang Ting , Li Jinggang
Published: November 2022

In order to establish and predict the flow and pressure distribution of the reactor coolant system of SMR under steady-state operating conditions, so as to provide an important design input for the thermal-hydraulic design of the reactor. This study is based on the flow characteristics of the SMR with multiple flow paths and multiple nodes and hydraulics property, and two numerical methods are established for the hydraulic analysis of the circulation system. For the flow that can be simplified into the flow path and the node, the finite element analysis method is combined with the quasi-linear processing of the flow and the head loss to realize the flow distribution calculation of the circulation system; For the upper head bypass flow with a certain pressure distribution, the method of bisection is used to achieve calculation of such type flow. The research results show that the numerical method established can be used for the flow and pressure distribution of the reactor coolant system of SMR, and the numerical calculation results are in line with expectations.
Research on the Impact of Advanced Rule Design System on the Digitalization of Reactor Building Model
Authors: Wang Jin , Yang Hongyi
Published: November 2022

This paper proposes a more stringent method for customizing project rules. This method customizes the comprehensive management of the project and component reference database on the digital plant design platform based on general design codes, standards, item classification principles in nuclear engineering, digitalization requirements in reactor design, plant layout, project management, and material procurement and construction, etc. To improve the correlation between design specifications and digital power plants, suggest enhancing the data consistency among different design disciplines. Standardize the three-dimensional layout design of nuclear power plants, and ensure the consistency between the digital power plant model and the natural power plant. The rules are sorted out, analyzed, and transformed systematically in this paper. These rules include the naming and classification principles of items in nuclear power projects, model data composition structure, essential attribute content, component selection filters, material performance, model parameters, output content format, vital requirements in plant layout for reactor design, etc. This paper will finally form a systematic rule customization scheme through refinement and improvement. The parameters, such as process, operating conditions, materials, fluids, specifications, safety, quality assurance, seismic and radioactivity levels, as well as item naming rules, project database, component reference database, three-dimensional modeling, information integration, attribute inheritance, data extraction, and other regulations. This scheme can make the three-dimensional arrangement more standard, the operation steps more concise, and significantly reduce the designer’s attribute range of manual input. It can effectively promote the accurate and appropriate expression of the process and instrumentation design scheme in the reactor building model. Significantly shorten the project design cycle. Data integration and transmission between rules enable system attributes to be deeply inherited and automatic checking and judging of operating conditions parameters and pressure and temperature limits in physical properties of component materials. This scheme can make the three-dimensional layout more standard, the operation steps more concise, significantly reduce the attributes range of manual input by the designer, and obtain better application feedback in automatic drawing and material reports. These rules provide more comprehensive data support for coupling experiments, data integration, process simulation, and digital handover of different disciplines and depths.
Study on Numerical Simulation Method of Unsteady Deposition Characteristics of Particles in Surface Filtration
Authors: Ma Song , Sun Zhongning , Zhou Yanmin , Gu Haifeng , Li Lintai
Published: November 2022

As an important part of the ventilation system of nuclear facilities, the performance of filter is of great significance to the environment and personnel safety. With the operation of the filter, the filtration mode will change from depth filtration to surface filtration. In this paper, a numerical simulation method is developed for the unsteady deposition characteristics of particles in the surface filtration. The parcel injection model is applied to accelerate the formation and growth process of dust cake on the surface of fiber filter. A series of subroutines running in FLUENT environment are developed to enhance the capabilities of computational fluid dynamics code in ANSYS. The Brownian force is added to the DPM model to calculate the particle trajectory more accurately. The function that aerosol particles are intercepted and deposited on the surface of fiber filter to form dust cake is added. The porous media model is used to simulate the fiber filter and dust cake. The porosity and resistance coefficient change with the unsteady deposition of particles. Combined with the subroutine with FLUENT software, a numerical simulation method which can be used to predict the unsteady deposition characteristics of particles in the surface filtration is formed. The numerical simulation results are compared with the experimental results, and they are basically consistent. So the developed numerical simulation method can be used to predict the unsteady deposition characteristics of particles in the surface filtration, and can also provide a theoretical basis for the structural design of pleated filter.
Development and Application of a Support Device for Checking Leakage of Heat Exchange Tubes of Spring Base Supporting Condenser in the Form of Filling Water
Authors: Xie Jianning
Published: November 2022

The installation and removal workload of the original water filling support of the condenser is huge. For example, the water filling leak detection of the condenser during outage takes a lot of time on the main line of the conventional island, which cannot meet the requirements of the outage period. Moreover, the support can only be removed after the condensate side of the condenser is filled with water and emptied, which needs to be further optimized. Based on the stress analysis and load modeling calculation of the original water-filled support and its legs, the stress distribution and deformation of the water-filled leak detection support of the condenser are determined, and the design is improved according to the demonstration results. Through rapid fixed function of self-locking wedge device design, design of jack-up device and synchronization automated design of hydraulic equipment, a new type of spring water filling pedestal supporting type condenser leakage inspection support device was developed, and thus the condenser water filling leak quick installation and dismantling becomes convenient, safe and reliable, which significantly shortens the condenser leakage in the process of water filling to check support operation time. At the same time, the synchronous lifting of the water-filled leak detection support ensures the structural safety of the condenser during the process of condenser descent when the condensate side of the condenser is drained to the operating level.
Study on Passive Pulse Cooling Method of Secondary Side in PWR Nuclear Power Plant
Authors: Wu Zhenhua , Xu Junjun , Wang Weijin , Li Runcheng , Tang Qi
Published: November 2022

Currently, the supplementary strategy of PWR nuclear power unit to deal with the Station Black-Out (SBO) and Total Loss of Feed-water (TLFW) accidents mainly depends on mobile facilities and manual operation. Under the SBO together with TLFW (SBO+TLFW) beyond design basis accident, the self-sustaining time after the loss of active accident mitigation measures (safety injection / auxiliary feed-water / containment spray) of the nuclear power plant is limited, and the operation time left to the emergency team is relatively less. Especially in the case of external disasters such as super typhoon, the reachability of temporary facilities and personnel is challenged. By analyzing the available resources of the secondary side of the unit after SBO together with TLFW accident, this paper puts forward the passive pulse cooling method of the secondary side: the steam generators are divided into power SG(s) and cooling SG(s), the high-pressure steam of the power SG(s) is used as the passive power source, inject the high-temperature deaerated water stored in the deaerator into the released cooling SG(s), and the core is cooled by pulse water injection and intermittent exhaust. Through qualitative analysis of cooling mechanism, operation strategy and operation mode are given to put into services, and the simulation machine and CATHARE model are used for verification test and simulation calculation respectively. The results show that the two calculation trends are basically the same, which can delay core melting time more than 7 hours after SBO+TLFW accident. According to the probabilistic safety analysis (PSA) of a plant, the core damage frequency (CDF) is reduced by about 7%. The research results in this paper will be helpful to optimize TLFW response strategies and avoid putting into “Feed and bleed mode” too early and contribute to provide mitigation measures for delaying core melting under SBO+TLFW accident, and to gain time for putting external mitigation measures into service.
Modeling and Fault Diagnosis of Pressurizer Based on Bond Graph
Authors: Zhou Zhuoran , Peng Minjun , Wang Hang , Jiang Yingying
Published: November 2022

The bond graph can be used to simulate and model conservation processes in different energy domains. In this paper, based on bond graph, the pressurizer was modeled and diagnosed at full power. Two parameters of pressure and mass flow were selected. Firstly, the bond graph of pressurizer was established. The pressurizer was divided into three zones: the fluctuating zone, the saturated liquid zone and the gaseous zone. The conservation equations of mass and energy were given respectively for these three regions. Secondly, the conservation equations of mass and energy were combined with the conservation of volume, and the model was constructed according to the combination rules. Then, according to the established bond graph model, the time causality diagram was established to determine the influence relationship between the coefficients and listed the fault characteristic matrix. Meanwhile, the reasoning structure between the coefficient was realized by the expert system G2. Finally, the fault data simulated by simulator was imported into the bond graph model. The failure of the safety valve, the spray valve and the immersion type heater were introduced to verify the modeling and diagnosis effect. It concludes that the diagnosis effect is consistent with the imported fault. Compared with the data-driven method, the bond graph method is more reliable, and compared with the expert knowledge method, the bond graph method can diagnose unknown faults and has a wide range of applications. In addition, a new method for nuclear power plant simulation modeling was provided. This was a new attempt to model bond graphs outside the research of mechanic, electronic and hydraulic.
Selection of Complex Accident Sequences for Sodium-Cooled Fast Reactor Design Extension Conditions
Authors: Jiang Jingke , Zhao Nanfei , Yang Dandan
Published: November 2022

Design Extension Conditions (DEC) are accident conditions beyond design basis for nuclear power plants, including complex accident sequences (DEC-A) and severe accidents (DEC-B). As required by the nuclear safety regulation HAF102-2016, DEC must be considered by identifying additional accident scenarios and develop practical prevention and mitigation measures for such accidents in design process for nuclear power plants. The aim is to enhance nuclear power plant’s response to beyond design basis accidents or multi-fault accident. It’s to contribute to increase the design requirements and expand the design consideration for nuclear power plants to consider DEC in design process, which may increase the design difficulty and construction and operation costs for nuclear power plants. The selection of DEC is great significance to balance the safety and economy for nuclear power plants.

Compared with DEC-B, the DEC-A is a newer concept, and the mature cognition of DEC-A has not yet been reached. The selection method of DEC-A for sodium-cooled fast reactor is analyzed systematically, which includes how to define, identify and screen DEC-A. Taking a sodium-cooled fast reactor as an example, DEC-A are identified and screened in the method of the combination of probabilistic safety assessment, determinism and engineering experience, and the entire selection process of DEC-A is performed. Besides, It is analyzed and discussed about the application of DEC-A in safety analysis and safety design of sodium-cooled fast reactors. Existing study results will provide theoretical reference for determining DEC-A lists for similar reactors.
Experiment Study on the Effect of Nozzle Shape on Liquid Jet Breakup
Authors: Sun Guofu , Zhan Yi , Okawa Tomio , Aoyagi Mitsuhiro , Uchibori Akihiro , Okano Yasushi
Published: November 2022

When a liquid sodium leakage accident occurs in a sodium-cooled fast reactor, the injected sodium collides with structures to produce splashing droplets, which can result in a violent combustion. According to previous studies on circular nozzles, the amount of splash is affected by the state of the jet at the moment of impact. However, the outlet shape of damaged area is hardly to be circular; and meanwhile it influences the flow pattern of jet a lot. Considering about this, in the present work, high-speed cameras were used to observe the jet discharged from oval nozzles vertically downward to investigate the falling process of the jet. The result shows that surface wave appears on the jet and within a certain range of flow velocity it can be observed obviously, meanwhile accelerate the breakup of jet.
A Proof-of-Concept Application of Fast Fracture Evaluation for the Primary Loop Section in Nuclear Power Plant
Authors: Zheng Zihao , Qiu Wenxin , Dai Yang , Min Jiesheng
Published: November 2022

The primary loop in nuclear power plants (NPPs) consists of piping and pressure vessel, which operate under a high level of pressure and temperature. During the long-term service, these sections need to have their integrity analyzed to check whether they might be prone to degradation, especially when exposed to radiation. In-service codes and standards such as the RSE-M and the ASME Section VIII enclose rules for the analysis of defects in piping systems. Although geometries for pipes are well-known and the fracture analyses methods are all today quite mature, it still takes an amount of time for structural engineers to rebuild models of different defects on piping section and the evaluation results varies with different modelling methods.

Hence, this paper provides a proof-of-concept tool for fast fracture evaluation, specific to the brittle fracture induced by thermal fatigue and radiation in the primary loop. In this application, the procedure of finite element method (FEM) analysis and fracture-related criteria is well integrated, based on the open-source FEM code code_aster and multi-tool platform SALOME. It automates several steps of primary loop section simulation, including mesh generation, calculation file configuration, post-processing and reporting results.

Furthermore, the verification and validation tests are built to ensure the robustness and reliability of the software and productivity.
Maintenance Strategy Optimization Based on AP-913 in Relay Protection Field
Authors: Zhao Yu , Li Ping , Tian Kunpeng , Fei Xingwei
Published: November 2022

AP-913 (Equipment Reliability Process Description) describes an equipment reliability process offered to assist member utilities to maintain high levels of safe and reliable plant operation in an efficient manner. This document based on AP-913 process document, this paper generally describes its determination by logic and analysis method. Combining with the types of failure and consequences in the actual problems, it created a more simplified maintenance strategy optimization mind maps (including the system stability, the design necessity, the economic efficiency, the equipment reliability four factors, refining the equipment and benefit related factors, its relationship and influence between the four are clarified), that can raise efficiency of maintenance and enhance the stability of the system. Through its application in the maintenance cases, it takes some cases for example: Increase the PM project of replacing the power board card of the transformer group, shorten the PM project cycle of the transformer group power transmitter, and change the control operation mode of the main transformer cooler;cancel the design of outdated manual working frequency excitation equipment and its PM items, cancellation of electric disk cabinet timer, etc. It is proved that the reliability of the equipment is improved after the implement of the maintenance strategy optimization. Time costs and labor costs are saved, and the cost of equipment procurement is decreased too. Responding to control the company’s cost positively, some ideas at the same time also can provide reference for other similar plant maintenance work.
PSA Application for UK HPR1000 ALARP Demonstration
Authors: Wang Zhiwen , Yang Chunju , Zhang Bing , Wang Jinkai , Sang Huina , Xiao Lingmei , Wang Fei , Wang Ming
Published: November 2022

UK version of the Hua-long Pressurised Reactor (UK HPR1000), designed by China General Nuclear Power Corporation (CGN), has been successfully completed the Generic Design Assessment (GDA), with the issue of a Design Acceptance Confirmation (DAC) from Office for Nuclear Regulation (ONR) and a Statement of Design Acceptability (SoDA) from the Environment Agency on 7 February 2022. In the UK GDA process, one of the main fundamental requirements is that the nuclear safety risk should be reduced to ALARP (As Low As Reasonably Practicable). During the UK HPR1000 design process, Probabilistic Safety Assessment (PSA) has been widely used as a key tool to inform the design and evaluate risk levels, and to demonstrate that the nuclear safety risk has been reduced to ALARP.

The process of PSA ALARP demonstration is established in accordance with the methodology that reflects the UK context requirements and consistent with relevant good practice. The process is fully applied by using PSA to systematically review the risk insights and to support other topic areas. After the systematic review of PSA risk insights from internal events Level 1 PSA, internal fire Level 1 PSA, internal flooding Level 1 PSA, external hazards Level 1PSA, external flooding Level 1 PSA, seismic PSA, Spent fuel pool PSA and Level 2 PSA, the related conservatisms, potential design improvements have been identified and addressed.

Current UK HPR1000 PSA results and risk insights show that the risk level of UK HPR1000 design is relatively low and the risk of UK HPR1000 design has been reduced to ALARP in GDA stage.