Cross-sectional homogenization for full-core calculations of small and complex reactor configurations,such as research reactors,has been recently recognized as an interesting and challenging topic.This paper presents ...Cross-sectional homogenization for full-core calculations of small and complex reactor configurations,such as research reactors,has been recently recognized as an interesting and challenging topic.This paper presents the development of a PARCS/Serpent model for the neutronics analysis of a research reactor type TRIGA Mark-II loaded with Russian VVR-M2 fuel(known as the Dalat Nuclear Research Reactor or DNRR).The full-scale DNRR model and a supercell model for a shim/safety rod and its surrounding fuel bundles with the Monte Carlo code Serpent 2 were proposed to generate homogenized fewgroup cross sections for full-core diffusion calculations with PARCS.The full-scale DNRR model with Serpent 2 was also utilized as a reference to verify the PARCS/Serpent calculations.Comparison of the effective neutron multiplication factors,radial and axial core power distributions,and control rod worths showed a generally good agreement between PARCS and Serpent 2.In addition,the discrepancies between the PARCS and Serpent 2 results are also discussed.Consequently,the results indicate the applicability of the PARCS/Serpent model for further steady state and transient analyses of the DNRR.展开更多
The determination of maintenance mode of complex equipment in nuclear power plant is an essential work for reliability analysis and maintenance decision. Currently, the main decision method of maintenance mode is reli...The determination of maintenance mode of complex equipment in nuclear power plant is an essential work for reliability analysis and maintenance decision. Currently, the main decision method of maintenance mode is reliability centered maintenance( RCM) logic decision-making process, but the process is a qualitative analysis process. Based on a comprehensive analysis of factors affecting equipment reliability and maintenance work, it adopts a fuzzy synthesis decision method to establish a maintenance decision model,which uses the maximum subordination principle and expert assessment method to determine the maintenance mode of complex equipment. Combined with a concrete example of generators in nuclear power plant,a description of maintenance decision method was proposed in the application of complex equipment. The research shows that the method is feasible and reliable.展开更多
An monitoring and earlywarning system is proposed for marine organisms and the cause of water intake blockage is analyzed. Based onthe intelligent sensing technology, computer software and hardware technology and digi...An monitoring and earlywarning system is proposed for marine organisms and the cause of water intake blockage is analyzed. Based onthe intelligent sensing technology, computer software and hardware technology and digital signal processing technology, the buoy monitoring platformsystem is developed by internet of things technology, cloud computing and the application of large data. Remote realtime monitoring of aquatic organisms and foreign bodies is realized based on underwater acoustic detection and low light imaging technology. Data processing center is established to store, analyze and process monitoring information and display it in real time, and provide emergency decision support. Through development and test of relevant key equipments, the reliability of cold source system of nuclear power plants is improved, which effectively reduces theinfluence of marine biological invasion on security and economic operation of the units.展开更多
Safe emplacement of high-level nuclear waste(HLNW)arising from the utilization of nuclear power is a frequently en-countered and considerably challenging issue.The widely accepted and feasible approach for the permane...Safe emplacement of high-level nuclear waste(HLNW)arising from the utilization of nuclear power is a frequently en-countered and considerably challenging issue.The widely accepted and feasible approach for the permanent disposal of HLNW involves housing it in a corrosion-resistant container and subsequently burying it deep in a geologic repository.The focus lies on ensuring the dur-ability and integrity of the container in this process.This review introduces various techniques and strategies employed in controlling the corrosion of used fuel containers(UFCs)using copper(Cu)as corrosion barrier in the context of deep geological disposal.Overall,these corrosion prevention techniques and methods have been effectively implemented and employed to successfully mitigate the corrosion challenges encountered during the permanent disposal of Cu containers(e.g.,corrosion mechanisms and corrosion parameters)in deep geologic repositories.The primary objective of this review is to provide an extensive examination of the alteration in the corrosion envir-onment encountered by the UFCs when subjected to deep geologic repository conditions and focusing on addressing the potential corro-sion scenarios.展开更多
This study aims to thoroughly investigate the axial power peaking factors (PPF) within the low-enriched uranium (LEU) core of the Ghana Research Reactor-1 (GHARR-1). This study uses advanced simulation tools, like the...This study aims to thoroughly investigate the axial power peaking factors (PPF) within the low-enriched uranium (LEU) core of the Ghana Research Reactor-1 (GHARR-1). This study uses advanced simulation tools, like the MCNPX code for analysing neutron behavior and the PARET/ANL code for understanding power variations, to get a clearer picture of the reactor’s performance. The analysis covers the initial six years of GHARR-1’s operation and includes projections for its whole 60-year lifespan. We closely observed the patterns of both the highest and average PPFs at 21 axial nodes, with measurements taken every ten years. The findings of this study reveal important patterns in power distribution within the core, which are essential for improving the safety regulations and fuel management techniques of the reactor. We provide a meticulous approach, extensive data, and an analysis of the findings, highlighting the significance of continuous monitoring and analysis for proactive management of nuclear reactors. The findings of this study not only enhance our comprehension of nuclear reactor safety but also carry significant ramifications for sustainable energy progress in Ghana and the wider global context. Nuclear engineering is essential in tackling global concerns, such as the demand for clean and dependable energy sources. Research on optimising nuclear reactors, particularly in terms of safety and efficiency, is crucial for the ongoing advancement and acceptance of nuclear energy.展开更多
The current research of Charpy impact mainly focuses on obtaining the ductile brittle transition temperature of materials by experiments.Compared with experiments,numerical simulation can study many problems with hars...The current research of Charpy impact mainly focuses on obtaining the ductile brittle transition temperature of materials by experiments.Compared with experiments,numerical simulation can study many problems with harsh conditions.However,there are still few studies on the influence of geometric factors such as side grooves.In this paper,the geometry of standard Charpy impact test is designed.Specimens with different widths and side grooves are tested.The finite element model of Charpy impact was established by ABAQUS software.Use test results and simulation results to verify each other.The effects of sample width,side groove depth and side groove bottom fillet on the impact fracture resistance of the sample were studied.The results show that the specimen width is positively correlated with the impact toughness of the specimen.The side groove greatly reduces the impact toughness of the material;the toughness of side groove decreases with the increase of depth;the fracture toughness of side groove decreases with the increase of fillet at the bottom of side groove.The proportion of toughness energy to impact energy of samples was analyzed.The results show that the toughness energy accounts for about 70%of the impact energy of the sample,which has little to do with the geometric characteristics of the sample.This study presents a reliable method for studying Charpy impact tests.The influence of geometric parameters is obtained,which provides a reference method for the study of impact toughness of high toughness materials.展开更多
As users’access to the network has evolved into the acquisition of mass contents instead of IP addresses,the IP network architecture based on end-to-end communication cannot meet users’needs.Therefore,the Informatio...As users’access to the network has evolved into the acquisition of mass contents instead of IP addresses,the IP network architecture based on end-to-end communication cannot meet users’needs.Therefore,the Information-Centric Networking(ICN)came into being.From a technical point of view,ICN is a promising future network architecture.Researching and customizing a reasonable pricing mechanism plays a positive role in promoting the deployment of ICN.The current research on ICN pricing mechanism is focused on paid content.Therefore,we study an ICN pricing model for free content,which uses game theory based on Nash equilibrium to analysis.In this work,advertisers are considered,and an advertiser model is established to describe the economic interaction between advertisers and ICN entities.This solution can formulate the best pricing strategy for all ICN entities and maximize the benefits of each entity.Our extensive analysis and numerical results show that the proposed pricing framework is significantly better than existing solutions when it comes to free content.展开更多
At T-junctions, where hot and cold streams flowing in pipes join and mix, significant temperature fluctuations can be created in very close neighborhood of the pipe walls. The wall temperature fluctuations cause cycli...At T-junctions, where hot and cold streams flowing in pipes join and mix, significant temperature fluctuations can be created in very close neighborhood of the pipe walls. The wall temperature fluctuations cause cyclical thermal stresses which may induce fatigue cracking. Temperature fluctuation is of crucial importance in many engineering applications and especially in nuclear power plants. This is because the phenomenon leads to thermal fatigue and might subsequently result in failure of structural material. Therefore, the effects of temperature fluctuation in piping structure at mixing junctions in nuclear power systems cannot be neglected. In nuclear power plant, piping structure is exposed to unavoidable temperature differences in a bid to maintain plant operational capacity. Tightly coupled to temperature fluctuation is flow turbulence, which has attracted extensive attention and has been investigated worldwide since several decades. The focus of this study is to investigate the effects of injection pipe orientation on flow mixing and temperature fluctuation for fluid flow downstream a T-junction. Computational fluid dynamics (CFD) approach was applied using STAR CCM+ code. Four inclination angles including 0 (90), 15, 30 and 45 degrees were studied and the mixing intensity and effective mixing zone were investigated. K-omega SST turbulence model was adopted for the simulations. Results of the analysis suggest that, effective mixing of cold and hot fluid which leads to reduced and uniform temperature field at the pipe wall boundary, is achieved at 0 (90) degree inclination of the branch pipe and hence may lower thermal stress levels in the structural material of the pipe. Turbulence mixing, pressure drop and velocity distribution were also found to be more appreciable at 0 (90) degree inclination angle of the branch pipe relative to the other orientations studied.展开更多
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
Taking the project of introducing reliability-centered maintenance( RCM) into maintenance decision in an AP1000 nuclear power plant( NPP) under construction as the research object,an improved RCM methodology was propo...Taking the project of introducing reliability-centered maintenance( RCM) into maintenance decision in an AP1000 nuclear power plant( NPP) under construction as the research object,an improved RCM methodology was proposed, and the application software and an RCM-based maintenance strategies management system were designed. In the pilot project,the RCMbased maintenance decision methodology had been applied to determining the maintenance strategies for two systems. Both the decision process and the results were described in this paper. The achievements of this project promoted the introduction and routinization of an advanced and effective maintenance decision mode in nuclear power field,which could provide valuable reference for new NPPs in China.展开更多
To introduce the basic concepts of technical specification of nuclear power plant,a risk assessment and management technique based on the probabilistic safety analysis( PSA) method was proposed. The risk-informed meth...To introduce the basic concepts of technical specification of nuclear power plant,a risk assessment and management technique based on the probabilistic safety analysis( PSA) method was proposed. The risk-informed method was used,and an example was given to show how to use some specific risk metrics like CDF / LERF /ICDP / ILERP to analyze and manage the risk associated with activities in nuclear power plant operation. The advantage of this technique can be concluded from this paper,and this technique should be used more widely and deeply in nuclear industry.展开更多
The microstructural evolution and Vickers hardness measurement in the welding heat-affected zone （HAZ） of HD15 Nil MnMoNbCu steel for nuclear power station were investigated by Gleeble-3180 thermal mechanical simula...The microstructural evolution and Vickers hardness measurement in the welding heat-affected zone （HAZ） of HD15 Nil MnMoNbCu steel for nuclear power station were investigated by Gleeble-3180 thermal mechanical simulator, and the simulated HAZ continuous cooling transformation curves （SH-CCT） were measured simultaneously. With ts/5 inereasing from 3.75 s to 15 000 s, the product was obtained martensite, bainite, ferrite and pearlite, successively. The result of microstructure and Vickers hardness in the heat-affected zone was in good agreement with those measured by SH-CCT diagram with the heat input 16. 2 kJ/cm as an example to weld the HD15Ni1MnMoNbCu steel pipe using TIG/SMAW/SAW welding methods.展开更多
A calculation method of heat transfer area for vertical natural circulated steam generator was introduced. According to the design requirements of steam generator 55/19 of CPR1000, its heat transfer area was calculate...A calculation method of heat transfer area for vertical natural circulated steam generator was introduced. According to the design requirements of steam generator 55/19 of CPR1000, its heat transfer area was calculated based on this method. The results show that the accuracy of partitional and overall calculation method is almost the same, but the result is different when using different calculation models. And the results are compared with the foreign companies for 55/19 steam generator.展开更多
A slightly prompt critical nuclear reactor would increase the neutron flux exponentially at a high rate causing the reactor to become uncontrollable, however due to the delayed neutrons, it is possible to leave the re...A slightly prompt critical nuclear reactor would increase the neutron flux exponentially at a high rate causing the reactor to become uncontrollable, however due to the delayed neutrons, it is possible to leave the reactor in a subcritical state as far as only prompt neutrons are concerned and to also sustain the chain reaction when it is going to die out. The delay neutron flux spectrum of the compact core of the Ghana’s miniature neutron source reactor (MNSR) was studied using the Monte Carlo method. 20,484 energy groups combined for all three categories of the energy distribution, thermal, slowing down and fast regions were modeled to create small energy bins. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the regions monitored. The delay thermal neutrons recorded its highest flux in the inner irradiation channel with an average flux of (4.0127 0.0076) × 1008 n/cm2?s, followed by the outer irradiation channel with an average flux of (2.4524 0.0049) × 1008 n/cm2?s. The beryllium reflector recorded the lowest flux in the thermal region. These values of the thermal energy range occurred in the energy range (0 – 0.625× 10 – 07) MeV. The inner irradiation channel again recorded the highest average flux of (1.2050 ± 0.0501) × 1007 n/cm2?s at the slowing down region in the energy range (0.821 – 6.94) MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast energy region, (6.96 – 20) MeV, the core, where the moderator is found, the same trend was observed with the inner irradiation channel recording the highest flux at an average flux of (2.0647 ± 0.3260) × 1006 n/cm2?s .The outer irradiation channel recorded the second highest flux while the annulus beryllium reflector recorded very low flux in this region. The final k-effective contribution from only delay neutrons is 0.00834 with the delay neutron fraction being 0.01357 ± 0.00049, hence the Ghana MNSR has good safety inherent feature.展开更多
External filtered neutron beams have been developed at the horizontal radial channels No. 4 of Dalat research reactor. In the material composition of the neutron filters, the primary material components of Iron, Alumi...External filtered neutron beams have been developed at the horizontal radial channels No. 4 of Dalat research reactor. In the material composition of the neutron filters, the primary material components of Iron, Aluminum, Nickel and Vanadium are used to obtain the mono-energetic neutron beams of 24 and 59 keV, with low level of Gamma and slow neutron background. A computer code and Monte-Carlo simulation technique were applied to optimize the filter configurations and to deduce the neutron energy distributions in the filtered beams. A hydrogen-filled proton recoil detector and the activation method with Gold foils were used to measure the neutron energy spectrum and flux of each beam at sample position. The results of experimental neutron fluxes are 6.1 × 105 and 5.3 × 105 n/cm2/s for 24 and 59 keV beams, respectively.展开更多
The National Nuclear Research Institute of the Ghana Atomic Energy Commission is undertaking steps to convert the Ghana Research Reactor-1 from HEU Core to LEU. The proposed LEU core consists of 12.5% enriched UO2 fue...The National Nuclear Research Institute of the Ghana Atomic Energy Commission is undertaking steps to convert the Ghana Research Reactor-1 from HEU Core to LEU. The proposed LEU core consists of 12.5% enriched UO2 fuel elements clad in Zircaloy-4 alloy. This is done in collaboration with Reduced Enrichment for Research and Test Reactor. The versatile MCNP code was used to analyse the neutronics parameters given in the SAR of HEU core, thereby characterizing the core. Subsequently, the LEU core was indentified with necessary changes to the HEU MCNP model. It was ascertained that the reactivity for the LEU core with the same number of fuel pins as the HEU was inadequate, hence the fuel pins were increased from 344 to 348. The neutron flux at the irradiation sites was found to be below the nominal value at full power for the LEU and hence the nominal power was increased to 34 kW for a nominal flux value of 1 × 1012 n/cm2.s. The parameters investigated for the HEU and LEU are shown in this paper.展开更多
With rapid development of nuclear power in China, in view of reactor type selection, this paper analyzes the current situation that faces nuclear power industry, the technical characteristics of optional reactors and ...With rapid development of nuclear power in China, in view of reactor type selection, this paper analyzes the current situation that faces nuclear power industry, the technical characteristics of optional reactors and the tendency of nuclear power technology development in the future. The proposals put forward in this paper include choosing and introducing GW-class advanced PWR as main reactors, carrying out self-supporting projects and technical transfer negotiations, in addition, promoting the design of the advanced generation-Ⅱ PWR and initiating small-scaled construction. The ultimate target is to catch up with the world advanced level by means of technical upgrading and recreation based on technology importation and assimilation.展开更多
The objective of this study is to design an elastic neutron scattering system<span><span style="font-family:;" "=""> according to the angle with a sample using thermal neutron beam...The objective of this study is to design an elastic neutron scattering system<span><span style="font-family:;" "=""> according to the angle with a sample using thermal neutron beam at the Dalat <span>Nuclear Reactor (DNR). The system is used for research and training in the</span> field <span>of material structure analysis by neutron scattering and diffraction tech</span>nique</span></span><span><span style="font-family:;" "="">s</span></span><span><span style="font-family:;" "="">. It is designed on the basis of inheriting the neutron measurement spectrometer systems at the DNR and the scattered neutron measurement systems in the world. The measuring system, which was installed at the hori<span>zontal channel</span></span></span><span><span style="font-family:;" "=""> </span></span><span><span style="font-family:;" "="">4 of the DNR, consists of </span></span><span><span style="font-family:;" "="">5-helium-3 detectors and a fully</span></span><span><span style="font-family:;" "=""> electronic system to record the scatter counts <span>and a mechanical system with the possibility of rotating at 15</span></span></span><span><span style="font-family:;" "=""><span style="white-space:nowrap;"><span style="white-space:nowrap;">˚</span></span></span></span><span><span style="font-family:;" "="">-</span></span><span><span style="font-family:;" "="">75</span></span><span><span style="font-family:;" "=""><span style="white-space:nowrap;"><span style="white-space:nowrap;">˚</span></span></span></span><span><span style="font-family:;" "=""> </span></span><span><span style="font-family:;" "="">angles. The constructed system is tested for <span>evaluation of the accuracy, stability and reliability of the mechanical and</span> electronic systems of moving detector</span></span><span><span style="font-family:;" "="">s</span></span><span><span style="font-family:;" "=""> by angles.</span></span>展开更多
基金the Ministry of Science and Technology of Vietnam(No.DTCB.06/18/VKHKTHN).
文摘Cross-sectional homogenization for full-core calculations of small and complex reactor configurations,such as research reactors,has been recently recognized as an interesting and challenging topic.This paper presents the development of a PARCS/Serpent model for the neutronics analysis of a research reactor type TRIGA Mark-II loaded with Russian VVR-M2 fuel(known as the Dalat Nuclear Research Reactor or DNRR).The full-scale DNRR model and a supercell model for a shim/safety rod and its surrounding fuel bundles with the Monte Carlo code Serpent 2 were proposed to generate homogenized fewgroup cross sections for full-core diffusion calculations with PARCS.The full-scale DNRR model with Serpent 2 was also utilized as a reference to verify the PARCS/Serpent calculations.Comparison of the effective neutron multiplication factors,radial and axial core power distributions,and control rod worths showed a generally good agreement between PARCS and Serpent 2.In addition,the discrepancies between the PARCS and Serpent 2 results are also discussed.Consequently,the results indicate the applicability of the PARCS/Serpent model for further steady state and transient analyses of the DNRR.
文摘The determination of maintenance mode of complex equipment in nuclear power plant is an essential work for reliability analysis and maintenance decision. Currently, the main decision method of maintenance mode is reliability centered maintenance( RCM) logic decision-making process, but the process is a qualitative analysis process. Based on a comprehensive analysis of factors affecting equipment reliability and maintenance work, it adopts a fuzzy synthesis decision method to establish a maintenance decision model,which uses the maximum subordination principle and expert assessment method to determine the maintenance mode of complex equipment. Combined with a concrete example of generators in nuclear power plant,a description of maintenance decision method was proposed in the application of complex equipment. The research shows that the method is feasible and reliable.
文摘An monitoring and earlywarning system is proposed for marine organisms and the cause of water intake blockage is analyzed. Based onthe intelligent sensing technology, computer software and hardware technology and digital signal processing technology, the buoy monitoring platformsystem is developed by internet of things technology, cloud computing and the application of large data. Remote realtime monitoring of aquatic organisms and foreign bodies is realized based on underwater acoustic detection and low light imaging technology. Data processing center is established to store, analyze and process monitoring information and display it in real time, and provide emergency decision support. Through development and test of relevant key equipments, the reliability of cold source system of nuclear power plants is improved, which effectively reduces theinfluence of marine biological invasion on security and economic operation of the units.
基金study received financial support from the National Natural Science Foundation of China(No.U22B2065),EditChecks(https://editchecks.com.cn/)for providing linguistic assistance during the preparation of this manuscript.
文摘Safe emplacement of high-level nuclear waste(HLNW)arising from the utilization of nuclear power is a frequently en-countered and considerably challenging issue.The widely accepted and feasible approach for the permanent disposal of HLNW involves housing it in a corrosion-resistant container and subsequently burying it deep in a geologic repository.The focus lies on ensuring the dur-ability and integrity of the container in this process.This review introduces various techniques and strategies employed in controlling the corrosion of used fuel containers(UFCs)using copper(Cu)as corrosion barrier in the context of deep geological disposal.Overall,these corrosion prevention techniques and methods have been effectively implemented and employed to successfully mitigate the corrosion challenges encountered during the permanent disposal of Cu containers(e.g.,corrosion mechanisms and corrosion parameters)in deep geologic repositories.The primary objective of this review is to provide an extensive examination of the alteration in the corrosion envir-onment encountered by the UFCs when subjected to deep geologic repository conditions and focusing on addressing the potential corro-sion scenarios.
文摘This study aims to thoroughly investigate the axial power peaking factors (PPF) within the low-enriched uranium (LEU) core of the Ghana Research Reactor-1 (GHARR-1). This study uses advanced simulation tools, like the MCNPX code for analysing neutron behavior and the PARET/ANL code for understanding power variations, to get a clearer picture of the reactor’s performance. The analysis covers the initial six years of GHARR-1’s operation and includes projections for its whole 60-year lifespan. We closely observed the patterns of both the highest and average PPFs at 21 axial nodes, with measurements taken every ten years. The findings of this study reveal important patterns in power distribution within the core, which are essential for improving the safety regulations and fuel management techniques of the reactor. We provide a meticulous approach, extensive data, and an analysis of the findings, highlighting the significance of continuous monitoring and analysis for proactive management of nuclear reactors. The findings of this study not only enhance our comprehension of nuclear reactor safety but also carry significant ramifications for sustainable energy progress in Ghana and the wider global context. Nuclear engineering is essential in tackling global concerns, such as the demand for clean and dependable energy sources. Research on optimising nuclear reactors, particularly in terms of safety and efficiency, is crucial for the ongoing advancement and acceptance of nuclear energy.
基金Supported by National Natural Science Foundation of China(Grant Nos.51975526,51505425)National Key R&D Program of China(Grant No.2018YFC0808800)+1 种基金Open Project of Key Laboratory of MEM of China(Grant No.2020XFZB10)Technical Service Projects(Grant Nos.HZFS-XZ-2022-07-02,XJBY-20211221).
文摘The current research of Charpy impact mainly focuses on obtaining the ductile brittle transition temperature of materials by experiments.Compared with experiments,numerical simulation can study many problems with harsh conditions.However,there are still few studies on the influence of geometric factors such as side grooves.In this paper,the geometry of standard Charpy impact test is designed.Specimens with different widths and side grooves are tested.The finite element model of Charpy impact was established by ABAQUS software.Use test results and simulation results to verify each other.The effects of sample width,side groove depth and side groove bottom fillet on the impact fracture resistance of the sample were studied.The results show that the specimen width is positively correlated with the impact toughness of the specimen.The side groove greatly reduces the impact toughness of the material;the toughness of side groove decreases with the increase of depth;the fracture toughness of side groove decreases with the increase of fillet at the bottom of side groove.The proportion of toughness energy to impact energy of samples was analyzed.The results show that the toughness energy accounts for about 70%of the impact energy of the sample,which has little to do with the geometric characteristics of the sample.This study presents a reliable method for studying Charpy impact tests.The influence of geometric parameters is obtained,which provides a reference method for the study of impact toughness of high toughness materials.
基金supported by the Key R&D Program of Anhui Province in 2020 under Grant No.202004a05020078China Environment for Network Innovations(CENI)under Grant No.2016-000052-73-01-000515.
文摘As users’access to the network has evolved into the acquisition of mass contents instead of IP addresses,the IP network architecture based on end-to-end communication cannot meet users’needs.Therefore,the Information-Centric Networking(ICN)came into being.From a technical point of view,ICN is a promising future network architecture.Researching and customizing a reasonable pricing mechanism plays a positive role in promoting the deployment of ICN.The current research on ICN pricing mechanism is focused on paid content.Therefore,we study an ICN pricing model for free content,which uses game theory based on Nash equilibrium to analysis.In this work,advertisers are considered,and an advertiser model is established to describe the economic interaction between advertisers and ICN entities.This solution can formulate the best pricing strategy for all ICN entities and maximize the benefits of each entity.Our extensive analysis and numerical results show that the proposed pricing framework is significantly better than existing solutions when it comes to free content.
文摘At T-junctions, where hot and cold streams flowing in pipes join and mix, significant temperature fluctuations can be created in very close neighborhood of the pipe walls. The wall temperature fluctuations cause cyclical thermal stresses which may induce fatigue cracking. Temperature fluctuation is of crucial importance in many engineering applications and especially in nuclear power plants. This is because the phenomenon leads to thermal fatigue and might subsequently result in failure of structural material. Therefore, the effects of temperature fluctuation in piping structure at mixing junctions in nuclear power systems cannot be neglected. In nuclear power plant, piping structure is exposed to unavoidable temperature differences in a bid to maintain plant operational capacity. Tightly coupled to temperature fluctuation is flow turbulence, which has attracted extensive attention and has been investigated worldwide since several decades. The focus of this study is to investigate the effects of injection pipe orientation on flow mixing and temperature fluctuation for fluid flow downstream a T-junction. Computational fluid dynamics (CFD) approach was applied using STAR CCM+ code. Four inclination angles including 0 (90), 15, 30 and 45 degrees were studied and the mixing intensity and effective mixing zone were investigated. K-omega SST turbulence model was adopted for the simulations. Results of the analysis suggest that, effective mixing of cold and hot fluid which leads to reduced and uniform temperature field at the pipe wall boundary, is achieved at 0 (90) degree inclination of the branch pipe and hence may lower thermal stress levels in the structural material of the pipe. Turbulence mixing, pressure drop and velocity distribution were also found to be more appreciable at 0 (90) degree inclination angle of the branch pipe relative to the other orientations studied.
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Taking the project of introducing reliability-centered maintenance( RCM) into maintenance decision in an AP1000 nuclear power plant( NPP) under construction as the research object,an improved RCM methodology was proposed, and the application software and an RCM-based maintenance strategies management system were designed. In the pilot project,the RCMbased maintenance decision methodology had been applied to determining the maintenance strategies for two systems. Both the decision process and the results were described in this paper. The achievements of this project promoted the introduction and routinization of an advanced and effective maintenance decision mode in nuclear power field,which could provide valuable reference for new NPPs in China.
文摘To introduce the basic concepts of technical specification of nuclear power plant,a risk assessment and management technique based on the probabilistic safety analysis( PSA) method was proposed. The risk-informed method was used,and an example was given to show how to use some specific risk metrics like CDF / LERF /ICDP / ILERP to analyze and manage the risk associated with activities in nuclear power plant operation. The advantage of this technique can be concluded from this paper,and this technique should be used more widely and deeply in nuclear industry.
文摘The microstructural evolution and Vickers hardness measurement in the welding heat-affected zone （HAZ） of HD15 Nil MnMoNbCu steel for nuclear power station were investigated by Gleeble-3180 thermal mechanical simulator, and the simulated HAZ continuous cooling transformation curves （SH-CCT） were measured simultaneously. With ts/5 inereasing from 3.75 s to 15 000 s, the product was obtained martensite, bainite, ferrite and pearlite, successively. The result of microstructure and Vickers hardness in the heat-affected zone was in good agreement with those measured by SH-CCT diagram with the heat input 16. 2 kJ/cm as an example to weld the HD15Ni1MnMoNbCu steel pipe using TIG/SMAW/SAW welding methods.
文摘A calculation method of heat transfer area for vertical natural circulated steam generator was introduced. According to the design requirements of steam generator 55/19 of CPR1000, its heat transfer area was calculated based on this method. The results show that the accuracy of partitional and overall calculation method is almost the same, but the result is different when using different calculation models. And the results are compared with the foreign companies for 55/19 steam generator.
文摘A slightly prompt critical nuclear reactor would increase the neutron flux exponentially at a high rate causing the reactor to become uncontrollable, however due to the delayed neutrons, it is possible to leave the reactor in a subcritical state as far as only prompt neutrons are concerned and to also sustain the chain reaction when it is going to die out. The delay neutron flux spectrum of the compact core of the Ghana’s miniature neutron source reactor (MNSR) was studied using the Monte Carlo method. 20,484 energy groups combined for all three categories of the energy distribution, thermal, slowing down and fast regions were modeled to create small energy bins. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the regions monitored. The delay thermal neutrons recorded its highest flux in the inner irradiation channel with an average flux of (4.0127 0.0076) × 1008 n/cm2?s, followed by the outer irradiation channel with an average flux of (2.4524 0.0049) × 1008 n/cm2?s. The beryllium reflector recorded the lowest flux in the thermal region. These values of the thermal energy range occurred in the energy range (0 – 0.625× 10 – 07) MeV. The inner irradiation channel again recorded the highest average flux of (1.2050 ± 0.0501) × 1007 n/cm2?s at the slowing down region in the energy range (0.821 – 6.94) MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast energy region, (6.96 – 20) MeV, the core, where the moderator is found, the same trend was observed with the inner irradiation channel recording the highest flux at an average flux of (2.0647 ± 0.3260) × 1006 n/cm2?s .The outer irradiation channel recorded the second highest flux while the annulus beryllium reflector recorded very low flux in this region. The final k-effective contribution from only delay neutrons is 0.00834 with the delay neutron fraction being 0.01357 ± 0.00049, hence the Ghana MNSR has good safety inherent feature.
文摘External filtered neutron beams have been developed at the horizontal radial channels No. 4 of Dalat research reactor. In the material composition of the neutron filters, the primary material components of Iron, Aluminum, Nickel and Vanadium are used to obtain the mono-energetic neutron beams of 24 and 59 keV, with low level of Gamma and slow neutron background. A computer code and Monte-Carlo simulation technique were applied to optimize the filter configurations and to deduce the neutron energy distributions in the filtered beams. A hydrogen-filled proton recoil detector and the activation method with Gold foils were used to measure the neutron energy spectrum and flux of each beam at sample position. The results of experimental neutron fluxes are 6.1 × 105 and 5.3 × 105 n/cm2/s for 24 and 59 keV beams, respectively.
文摘The National Nuclear Research Institute of the Ghana Atomic Energy Commission is undertaking steps to convert the Ghana Research Reactor-1 from HEU Core to LEU. The proposed LEU core consists of 12.5% enriched UO2 fuel elements clad in Zircaloy-4 alloy. This is done in collaboration with Reduced Enrichment for Research and Test Reactor. The versatile MCNP code was used to analyse the neutronics parameters given in the SAR of HEU core, thereby characterizing the core. Subsequently, the LEU core was indentified with necessary changes to the HEU MCNP model. It was ascertained that the reactivity for the LEU core with the same number of fuel pins as the HEU was inadequate, hence the fuel pins were increased from 344 to 348. The neutron flux at the irradiation sites was found to be below the nominal value at full power for the LEU and hence the nominal power was increased to 34 kW for a nominal flux value of 1 × 1012 n/cm2.s. The parameters investigated for the HEU and LEU are shown in this paper.
文摘With rapid development of nuclear power in China, in view of reactor type selection, this paper analyzes the current situation that faces nuclear power industry, the technical characteristics of optional reactors and the tendency of nuclear power technology development in the future. The proposals put forward in this paper include choosing and introducing GW-class advanced PWR as main reactors, carrying out self-supporting projects and technical transfer negotiations, in addition, promoting the design of the advanced generation-Ⅱ PWR and initiating small-scaled construction. The ultimate target is to catch up with the world advanced level by means of technical upgrading and recreation based on technology importation and assimilation.
文摘The objective of this study is to design an elastic neutron scattering system<span><span style="font-family:;" "=""> according to the angle with a sample using thermal neutron beam at the Dalat <span>Nuclear Reactor (DNR). The system is used for research and training in the</span> field <span>of material structure analysis by neutron scattering and diffraction tech</span>nique</span></span><span><span style="font-family:;" "="">s</span></span><span><span style="font-family:;" "="">. It is designed on the basis of inheriting the neutron measurement spectrometer systems at the DNR and the scattered neutron measurement systems in the world. The measuring system, which was installed at the hori<span>zontal channel</span></span></span><span><span style="font-family:;" "=""> </span></span><span><span style="font-family:;" "="">4 of the DNR, consists of </span></span><span><span style="font-family:;" "="">5-helium-3 detectors and a fully</span></span><span><span style="font-family:;" "=""> electronic system to record the scatter counts <span>and a mechanical system with the possibility of rotating at 15</span></span></span><span><span style="font-family:;" "=""><span style="white-space:nowrap;"><span style="white-space:nowrap;">˚</span></span></span></span><span><span style="font-family:;" "="">-</span></span><span><span style="font-family:;" "="">75</span></span><span><span style="font-family:;" "=""><span style="white-space:nowrap;"><span style="white-space:nowrap;">˚</span></span></span></span><span><span style="font-family:;" "=""> </span></span><span><span style="font-family:;" "="">angles. The constructed system is tested for <span>evaluation of the accuracy, stability and reliability of the mechanical and</span> electronic systems of moving detector</span></span><span><span style="font-family:;" "="">s</span></span><span><span style="font-family:;" "=""> by angles.</span></span>
