Additive manufacturing(AM)is an innovative technique that enables the flexible design and construction of three-dimensional objects.In the nuclear industry,AM enables the use of advanced materials and high-performance...Additive manufacturing(AM)is an innovative technique that enables the flexible design and construction of three-dimensional objects.In the nuclear industry,AM enables the use of advanced materials and high-performance components.Although AM processing has been extensively investigated,the corresponding mechanical properties and structural integrity issues of AM parts have received less attention.This study reviews the mechanical behavior and key challenges of typical AM materials,fuel components,compact heat exchangers with complex geometries,and additive repair of damaged reactor components.The findings of this review will guide the efficient and reliable implementation of AM techniques in nuclear reactors.展开更多
A floating nuclear power plant(FNPP)is an offshore facility that integrates proven light-water reactor technologies with floating platform characteristics.However,frequent contact with marine environments may lead to ...A floating nuclear power plant(FNPP)is an offshore facility that integrates proven light-water reactor technologies with floating platform characteristics.However,frequent contact with marine environments may lead to wave-induced vibrations and oscillations.This study aimed to evaluate the wave danger on FNPPs,which can negatively impact FNPP functionality.We developed a hydrodynamic model of an FNPP using potential flow theory and computed the frequency-domain fluid dynamic responses.After verifying the hydrodynamic model,we developed a predictive model for FNPP responses.This model utilizes a genetic aggregation methodology for batch prediction while ensuring accuracy.We analyzed all the wave data from a selected sea area over the past 50 years using the constructed surrogate model,enabling us to identify dangerous marine areas.By utilizing the extreme value distribution of important wave heights in these areas,we determined the wave return period,which poses a threat to FNPPs.This provides an important method for analyzing wave hazards to FNPPs.展开更多
The integration of additive manufacturing(AM)and topology optimization(TO)has revolutionized the design and production of advanced equipment,providing innovative approaches to solving complex engineering challenges.In...The integration of additive manufacturing(AM)and topology optimization(TO)has revolutionized the design and production of advanced equipment,providing innovative approaches to solving complex engineering challenges.In the nuclear energy sector,achieving an optimal balance between the thermal and hydraulic performance of prismatic fuel elements has long been a key challenge.This study utilizes a coupled fluid-thermal TO method to design fuel elements with one,three,five,and seven inlets/outlets configurations suitable for AM.We systematically examine the impact of varying the number of inlets/outlets on the thermal-hydraulic performance of the elements.The results show that increasing the number of inlets/outlets can enhance the thermal performance of the fuel elements while sacrificing the hydraulic performance.Compared with the conventional design,the 5 inlets/outlets configuration achieved a coordinated improvement in both thermal and hydraulic performance,with a 2.38%enhancement in thermal performance and a 4.38%improvement in hydraulic performance.These findings highlight the significant potential of TO in improving the performance of fuel elements and strongly demonstrate the advantages of the collaborative application of AM and TO.展开更多
In pressurized water reactor(PWR),fretting wear is one of the main causes of fuel assembly failure.Moreover,the operation condition of cladding is complex and harsh.A unique fretting damage test equipment was develope...In pressurized water reactor(PWR),fretting wear is one of the main causes of fuel assembly failure.Moreover,the operation condition of cladding is complex and harsh.A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR environment.It can simulate the fretting wear experiment of PWR under different temperatures(maximum temperature is 350℃),displacement amplitude,vibration frequency,and normal force.The fretting wear behavior of Zr-4 alloy under different temperature environments was tested.In addition,the evolution of wear scar morphology,profile,and wear volume was studied using an optical microscope(OM),scanning electron microscopy(SEM),and a 3D white light interferometer.Results show that higher water temperature evidently decreased the cladding wear volume,the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.展开更多
Localized deformation and corrosion in irradiated 304 nuclear grade stainless steel in simulated primary water were investigated.The investigation was conducted by comparing the deformation structure,the oxide scale f...Localized deformation and corrosion in irradiated 304 nuclear grade stainless steel in simulated primary water were investigated.The investigation was conducted by comparing the deformation structure,the oxide scale formed at the deformation structure,and their correlation with cracking.The results revealed that increasing the irradiation dose promoted localized corrosion at the slip step and grain boundary,which was primarily attributed to the strain concentration induced by enhanced localized deformation and depletion of Cr at grain boundary.Further,a synergic effect of the enhanced localized deformation and localized corrosion at the slip step and grain boundary caused a higher cracking susceptibility of the irradiated steel.展开更多
Accurate and efficient online parameter identification and state estimation are crucial for leveraging digital twin simulations to optimize the operation of near-carbon-free nuclear energy systems.In previous studies,...Accurate and efficient online parameter identification and state estimation are crucial for leveraging digital twin simulations to optimize the operation of near-carbon-free nuclear energy systems.In previous studies,we developed a reactor operation digital twin(RODT).However,non-differentiabilities and discontinuities arise when employing machine learning-based surrogate forward models,challenging traditional gradient-based inverse methods and their variants.This study investigated deterministic and metaheuristic algorithms and developed hybrid algorithms to address these issues.An efficient modular RODT software framework that incorporates these methods into its post-evaluation module is presented for comprehensive comparison.The methods were rigorously assessed based on convergence profiles,stability with respect to noise,and computational performance.The numerical results show that the hybrid KNNLHS algorithm excels in real-time online applications,balancing accuracy and efficiency with a prediction error rate of only 1%and processing times of less than 0.1 s.Contrastingly,algorithms such as FSA,DE,and ADE,although slightly slower(approximately 1 s),demonstrated higher accuracy with a 0.3%relative L_2 error,which advances RODT methodologies to harness machine learning and system modeling for improved reactor monitoring,systematic diagnosis of off-normal events,and lifetime management strategies.The developed modular software and novel optimization methods presented offer pathways to realize the full potential of RODT for transforming energy engineering practices.展开更多
A severe accident in a marine nuclear reactor leads to radionuclide leakage,which causes hidden dangers to workers and has adverse effects of environmental pollution.It is necessary to propose a novel approach to radi...A severe accident in a marine nuclear reactor leads to radionuclide leakage,which causes hidden dangers to workers and has adverse effects of environmental pollution.It is necessary to propose a novel approach to radionuclide diffusion in a confined environment after a severe accident in a marine nuclear reactor.Therefore,this study proposes a new method for the severe accident analysis program MELCOR coupled with computational fluid dynamics scSTREAM to study radioactive diffusion in severe accidents.The radionuclide release fraction and temperature calculated by MELCOR were combined with the scSTREAM calculations to study the radionuclide diffusion behavior and the phenomenon of radionuclide diffusion in different space environments of the reactor under the conditions of varying wind velocities of the ventilation system and diffusion speed.The results show that the wind velocity of the ventilation system is very small or zero,and the turbulent diffusion of radionuclides is not obvious and diffuses slowly in the form of condensation sedimentation and gravity settlement.When the wind speed of the ventilation system increases,the flow of radionuclides meets the wall and forms eddy currents,affecting the time variation of radionuclides diffusing into chamber 2.The wind velocity of the ventilation system and the diffusion speed has opposite effects on the time variation trend of radionuclide diffusion into the four chambers.展开更多
The current research of nuclear control rod drive mechanism(CRDM)movable latch only makes a simple measurement of wear mass.The wear volume and difference in various claw surfaces are ignored and the degradation mecha...The current research of nuclear control rod drive mechanism(CRDM)movable latch only makes a simple measurement of wear mass.The wear volume and difference in various claw surfaces are ignored and the degradation mechanism of each claw surface is not clear.In this paper,a detailed degradation analysis was carried out on each claw surface of movable latch combined with wear result and worn morphology.Results indicate that the boundary of carbide is preferred for corrosion because carbide presents a nobler Volta potential compared to the metal matrix or boundary region.Due to the oscillation of drive shaft between the claw surfaces of movable latch,the dominant wear mechanism on the upper surface of claw(USC)and lower surface of claw(LSC)is plastic deformation caused by impact wear.Mechanical impact wear will cause the fragmentation of carbides because of the high hardness and low ductility of carbides.Corrosion promotes the broken carbides to fall off from the metal matrix.The generated fine carbides(abrasive particles)cause extra abrasive wear on USC when the movable brings the drive shaft upward or downward.As a result,USC has a higher wear volume than LSC.This research proposes a method to evaluate the wear on the whole movable latches using a 3D full-size scanner.展开更多
Corrosion behavior and mechanism of irradiated 304 nuclear grade stainless steel were studied in simulated pressurized water reactor primary water.The microstructure of the oxide formed on the steel irradiated to diff...Corrosion behavior and mechanism of irradiated 304 nuclear grade stainless steel were studied in simulated pressurized water reactor primary water.The microstructure of the oxide formed on the steel irradiated to different doses over an exposure period range of 25–1500 h was analyzed and compared.It was found that the general and intergranular corrosion rates of the steel were increased with irradiation dose,in correspondence with an evolution of the general oxide and the oxide formed at the grain boundary.Correlation of the oxide evolution with the corrosion kinetics and mechanism has been discussed in detail.展开更多
Numerous irradiation-induced gas bubbles are created in the nuclear fuel during irradiation, leading to the change of microstructure and the degradation of mechanical and thermal properties. The grain size of fuel is ...Numerous irradiation-induced gas bubbles are created in the nuclear fuel during irradiation, leading to the change of microstructure and the degradation of mechanical and thermal properties. The grain size of fuel is one of the important factors affecting bubble evolution. In current study, we first predict the thermodynamic behaviors of point defects as well as the interplay between vacancy and gas atom in both UO_(2) and U_(3)Si_(2) according to ab initio approach. Then, we establish the irradiation-induced bubble phase-field model to investigate the formation and evolution of intra-and inter-granular gas bubbles. The effects of fission rate and temperature on the evolutions of bubble morphologies in UO_(2) and U_(3)Si_(2) have been revealed. Especially, a comparison of porosities under different grain sizes is examined and analyzed. To understand the thermal conductivity as functions of grain size and porosity, the heat transfer capability of U_(3)Si_(2) is evaluated.展开更多
This paper presents the ageing mechanism of fuse in nuclear power plant in detail. Metal Electromigration is identified as the dominant ageing mechanism. On this basis, the dominant status indicators, temperature and ...This paper presents the ageing mechanism of fuse in nuclear power plant in detail. Metal Electromigration is identified as the dominant ageing mechanism. On this basis, the dominant status indicators, temperature and resistance of fuse were ensured, and current-temperature curve was proposed. The infrared thermal imaging technology was used to inspect the ageing condition and prove the current-temperature curve. Finally, the accelerated ageing testing was conducted abiding by the dominant ageing mechanism, and the lifetime was evaluated.展开更多
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 field of nuclear agriculture is introduced very briefly. Nuclear agriculture research/development system in Pakistan is described highlighting the achievements of the system partners. Description and discussion ar...The field of nuclear agriculture is introduced very briefly. Nuclear agriculture research/development system in Pakistan is described highlighting the achievements of the system partners. Description and discussion are generalized in concluding remarks at the end of the article. This article is an experimental guide for a developing nuclear agriculture system.展开更多
FeCrAl alloy is one of the most promising candidates as an accident-tolerant fuel(ATF)cladding material.Herein,the influence of cold-rolling(CR)reduction on microstructure and tensile properties of the as-annealed FeC...FeCrAl alloy is one of the most promising candidates as an accident-tolerant fuel(ATF)cladding material.Herein,the influence of cold-rolling(CR)reduction on microstructure and tensile properties of the as-annealed FeCrAl alloys,with low Cr and Nb contents,is systematically examined.With the increase in CR reduction,the grain size of FeCrAl alloy is obviously refined after annealing because the increase in stored deformation energy leads to enhanced recrystallization.However,the large CR reductions result in a severe mixed-grain microstructure,significantly reducing the uniform deformability of the FeCrAl alloy.The dislocation density of the as-annealed FeCrAl alloy decreases with the increase in CR reduction,except for the excessive CR reduction of 50%.Moreover,the Laves phases are crushed and dissolved during CR and annealing,as well as large amounts of refined Laves phases are found after large CR reductions.The pinning effect of the Laves phases can significantly improve the strength of FeCrAl alloy.Accordingly,the strengthening mechanisms of FeCrAl alloy consist of fine-grain strengthening,dislocation strengthening and precipitation strengthening.Finally,the FeCrAl alloy,with a CR reduction of 30%,achieves optimal tensile properties.This study can provide theoretical guidance for the industrial production of the FeCrAl alloy.展开更多
In this work,neodymium(Nd)and samarium(Sm)codoped lead lanthanum zirconate titanate(PLZT)ceramics were prepared by a high-temperature solid-state method.The samples were characterized by X-ray diffraction,scanning ele...In this work,neodymium(Nd)and samarium(Sm)codoped lead lanthanum zirconate titanate(PLZT)ceramics were prepared by a high-temperature solid-state method.The samples were characterized by X-ray diffraction,scanning electron microscopy and ferroelectric analysis.Rare earth-doped PLZT ceramics show good phase formation.An appropriate rare earth element doping amount increases the densities of PLZT ceramics and reduces their resistivities,which is due to the role of rare earth elements in grain refinement.However,the increase in the amount of grain boundaries caused by grain refinement also affects domain inversion.Therefore,with increasing doping concentration,the remnant polarization of PLZT gradually decreases,and the doping of rare earth elements also slightly reduces the band gap of PLZT.Under irradiation with an X-ray simulated beta source with a particle energy of 10 keV(between the average energies of the beta particles of^3H and^(63)Ni),the ceramic sheets in this work produce current densities of up to 1.38 nA/cm^2.This indicates that Nd and Sm codoped PLZT ceramics have a certain potential for application in betavoltaic batteries.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
Computational simulation is an important technical means in research of nuclear fuel materials.Since nuclear fuel issues are inherently multi-scopic,it is imperative to study them with multi-scale simulation scheme.At...Computational simulation is an important technical means in research of nuclear fuel materials.Since nuclear fuel issues are inherently multi-scopic,it is imperative to study them with multi-scale simulation scheme.At present,the development of multi-scale simulation for nuclear fuel materials calls for a more systematic approach,in which lies the main purpose of this article.The most important thing in multi-scale simulation is to accurately formulate the goals to be achieved and the types of methods to be used.In this regard,we first summarize the basic principles and applicability of the simulation methods which are commonly used in nuclear fuel research and are based on different scales ranging from micro to macro,i.e.First-Principles(FP),Molecular Dynamics(MD),Kinetic Monte Carlo(KMC),Phase Field(PF),Rate Theory(RT),and Finite Element Method(FEM).And then we discuss the major material issues in this field,also ranging from micro-scale to macro-scale and covering both pellets and claddings,with emphasis on what simulation method would be most suitable for solving each of the issues.Finally,we give our prospective analysis and understanding about the feasible ways of multi-scale integration and relevant handicaps and challenges.展开更多
The high burn-up structure(HBS)is characterized by the grain size of 100-300 nm and a porosity of up to 20%,which is formed at the rim of the nuclear fuel pellet due to 2-3 times higher local burn-up during the in-pil...The high burn-up structure(HBS)is characterized by the grain size of 100-300 nm and a porosity of up to 20%,which is formed at the rim of the nuclear fuel pellet due to 2-3 times higher local burn-up during the in-pile irradiation.HBS is considered a new potential structure for high-performance fuels.However,it is difficult to prepare HBS by conventional sintering methods.In this study,flash sintering was used to prepare HBS using CeO_(2)as a surrogate for a preliminary investigation.A new experimental configuration for rapid sintering of CeO_(2)pellets was provided,in which the green body can be rapidly preheated and pressure-assisted by the induction heating electrodes.An insulated quartz tube was used as the die for the flash sintered samples,allowing the current to flow through the sample and providing a stable condition for applying an external pressure of approximately 5.3-7.0 MPa during flash sintering process.Using an initial electric field of 141 V cm-1 and holding for 1-7 min at the maximum current density of~98 mA mm^(-2),CeO_(2)ceramics with a grain size of 114-282 nm and a relative density of 75.4-99.7%were prepared.The densification and microstructure evolution behaviors during flash sintering in this new experimental configuration have been discussed in detail.This new experimental configuration may provide a promising approach for preparing UO_(2)ceramics and their HBS.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supported by Stable Support Initiative of the State Key Laboratory of Advanced Nuclear Energy Technology(Grant No.YNSW-0124-0101-03)the author Cheng Zhang would like to thank the support by Stable Support Program for Scientific Research of China(Grant No.WDZC-2023-02-02-05).
文摘Additive manufacturing(AM)is an innovative technique that enables the flexible design and construction of three-dimensional objects.In the nuclear industry,AM enables the use of advanced materials and high-performance components.Although AM processing has been extensively investigated,the corresponding mechanical properties and structural integrity issues of AM parts have received less attention.This study reviews the mechanical behavior and key challenges of typical AM materials,fuel components,compact heat exchangers with complex geometries,and additive repair of damaged reactor components.The findings of this review will guide the efficient and reliable implementation of AM techniques in nuclear reactors.
文摘A floating nuclear power plant(FNPP)is an offshore facility that integrates proven light-water reactor technologies with floating platform characteristics.However,frequent contact with marine environments may lead to wave-induced vibrations and oscillations.This study aimed to evaluate the wave danger on FNPPs,which can negatively impact FNPP functionality.We developed a hydrodynamic model of an FNPP using potential flow theory and computed the frequency-domain fluid dynamic responses.After verifying the hydrodynamic model,we developed a predictive model for FNPP responses.This model utilizes a genetic aggregation methodology for batch prediction while ensuring accuracy.We analyzed all the wave data from a selected sea area over the past 50 years using the constructed surrogate model,enabling us to identify dangerous marine areas.By utilizing the extreme value distribution of important wave heights in these areas,we determined the wave return period,which poses a threat to FNPPs.This provides an important method for analyzing wave hazards to FNPPs.
基金supported by National Natural Science Foundation of China(Grant No.1257021702)National Key Research and Development Program of China(Grant No.2022YFB4603101).
文摘The integration of additive manufacturing(AM)and topology optimization(TO)has revolutionized the design and production of advanced equipment,providing innovative approaches to solving complex engineering challenges.In the nuclear energy sector,achieving an optimal balance between the thermal and hydraulic performance of prismatic fuel elements has long been a key challenge.This study utilizes a coupled fluid-thermal TO method to design fuel elements with one,three,five,and seven inlets/outlets configurations suitable for AM.We systematically examine the impact of varying the number of inlets/outlets on the thermal-hydraulic performance of the elements.The results show that increasing the number of inlets/outlets can enhance the thermal performance of the fuel elements while sacrificing the hydraulic performance.Compared with the conventional design,the 5 inlets/outlets configuration achieved a coordinated improvement in both thermal and hydraulic performance,with a 2.38%enhancement in thermal performance and a 4.38%improvement in hydraulic performance.These findings highlight the significant potential of TO in improving the performance of fuel elements and strongly demonstrate the advantages of the collaborative application of AM and TO.
基金Supported by National Key R&D Program of China(Grant No.2022YFB3401901)Key Program of National Natural Science Foundation of China(Grant No.U2067221)+2 种基金Sichuan Provincial Science and Technology Planning Project(Grant Nos.2022JDJQ0019 and 2022ZYD0029)Funds for China Postdoctoral Science Foundation(Grant No.2022M713008)Sichuan Provincial Innovative Talent Funding Project for Postdoctoral Fellows(Grant No.BX202225).
文摘In pressurized water reactor(PWR),fretting wear is one of the main causes of fuel assembly failure.Moreover,the operation condition of cladding is complex and harsh.A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR environment.It can simulate the fretting wear experiment of PWR under different temperatures(maximum temperature is 350℃),displacement amplitude,vibration frequency,and normal force.The fretting wear behavior of Zr-4 alloy under different temperature environments was tested.In addition,the evolution of wear scar morphology,profile,and wear volume was studied using an optical microscope(OM),scanning electron microscopy(SEM),and a 3D white light interferometer.Results show that higher water temperature evidently decreased the cladding wear volume,the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.
基金financially supported by the International Science&Technology Cooperation Program of China(No.2014DFA50800)partly supported by the Essential Research Fund by SNPTC(No.2015SN010-007)support of Natural Science Foundation of Jiangsu Province(BK20191178)。
文摘Localized deformation and corrosion in irradiated 304 nuclear grade stainless steel in simulated primary water were investigated.The investigation was conducted by comparing the deformation structure,the oxide scale formed at the deformation structure,and their correlation with cracking.The results revealed that increasing the irradiation dose promoted localized corrosion at the slip step and grain boundary,which was primarily attributed to the strain concentration induced by enhanced localized deformation and depletion of Cr at grain boundary.Further,a synergic effect of the enhanced localized deformation and localized corrosion at the slip step and grain boundary caused a higher cracking susceptibility of the irradiated steel.
基金supported by the Natural Science Foundation of Shanghai(No.23ZR1429300)Innovation Funds of CNNC(Lingchuang Fund,Contract No.CNNC-LCKY-202234)the Project of the Nuclear Power Technology Innovation Center of Science Technology and Industry(No.HDLCXZX-2023-HD-039-02)。
文摘Accurate and efficient online parameter identification and state estimation are crucial for leveraging digital twin simulations to optimize the operation of near-carbon-free nuclear energy systems.In previous studies,we developed a reactor operation digital twin(RODT).However,non-differentiabilities and discontinuities arise when employing machine learning-based surrogate forward models,challenging traditional gradient-based inverse methods and their variants.This study investigated deterministic and metaheuristic algorithms and developed hybrid algorithms to address these issues.An efficient modular RODT software framework that incorporates these methods into its post-evaluation module is presented for comprehensive comparison.The methods were rigorously assessed based on convergence profiles,stability with respect to noise,and computational performance.The numerical results show that the hybrid KNNLHS algorithm excels in real-time online applications,balancing accuracy and efficiency with a prediction error rate of only 1%and processing times of less than 0.1 s.Contrastingly,algorithms such as FSA,DE,and ADE,although slightly slower(approximately 1 s),demonstrated higher accuracy with a 0.3%relative L_2 error,which advances RODT methodologies to harness machine learning and system modeling for improved reactor monitoring,systematic diagnosis of off-normal events,and lifetime management strategies.The developed modular software and novel optimization methods presented offer pathways to realize the full potential of RODT for transforming energy engineering practices.
基金supported by the Postgraduate Scientific Research Innovation Project of Hunan Province (No. CX20210922)
文摘A severe accident in a marine nuclear reactor leads to radionuclide leakage,which causes hidden dangers to workers and has adverse effects of environmental pollution.It is necessary to propose a novel approach to radionuclide diffusion in a confined environment after a severe accident in a marine nuclear reactor.Therefore,this study proposes a new method for the severe accident analysis program MELCOR coupled with computational fluid dynamics scSTREAM to study radioactive diffusion in severe accidents.The radionuclide release fraction and temperature calculated by MELCOR were combined with the scSTREAM calculations to study the radionuclide diffusion behavior and the phenomenon of radionuclide diffusion in different space environments of the reactor under the conditions of varying wind velocities of the ventilation system and diffusion speed.The results show that the wind velocity of the ventilation system is very small or zero,and the turbulent diffusion of radionuclides is not obvious and diffuses slowly in the form of condensation sedimentation and gravity settlement.When the wind speed of the ventilation system increases,the flow of radionuclides meets the wall and forms eddy currents,affecting the time variation of radionuclides diffusing into chamber 2.The wind velocity of the ventilation system and the diffusion speed has opposite effects on the time variation trend of radionuclide diffusion into the four chambers.
基金Supported by Sichuan Science and Technology Program(Grant No.2019ZDZX0001)National Natural Science Foundation of China(Grant No.U2067221)Sichuan Science and Technology Planning Project(Grant No.22JCQN0111).
文摘The current research of nuclear control rod drive mechanism(CRDM)movable latch only makes a simple measurement of wear mass.The wear volume and difference in various claw surfaces are ignored and the degradation mechanism of each claw surface is not clear.In this paper,a detailed degradation analysis was carried out on each claw surface of movable latch combined with wear result and worn morphology.Results indicate that the boundary of carbide is preferred for corrosion because carbide presents a nobler Volta potential compared to the metal matrix or boundary region.Due to the oscillation of drive shaft between the claw surfaces of movable latch,the dominant wear mechanism on the upper surface of claw(USC)and lower surface of claw(LSC)is plastic deformation caused by impact wear.Mechanical impact wear will cause the fragmentation of carbides because of the high hardness and low ductility of carbides.Corrosion promotes the broken carbides to fall off from the metal matrix.The generated fine carbides(abrasive particles)cause extra abrasive wear on USC when the movable brings the drive shaft upward or downward.As a result,USC has a higher wear volume than LSC.This research proposes a method to evaluate the wear on the whole movable latches using a 3D full-size scanner.
基金financially supported by the International Science&Technology Cooperation Program of China(No.2014DFA50800)partly supported by the Essential Research Fundby SNPTC(No.2015 SN010-007)。
文摘Corrosion behavior and mechanism of irradiated 304 nuclear grade stainless steel were studied in simulated pressurized water reactor primary water.The microstructure of the oxide formed on the steel irradiated to different doses over an exposure period range of 25–1500 h was analyzed and compared.It was found that the general and intergranular corrosion rates of the steel were increased with irradiation dose,in correspondence with an evolution of the general oxide and the oxide formed at the grain boundary.Correlation of the oxide evolution with the corrosion kinetics and mechanism has been discussed in detail.
基金Project supported by the National Natural Science Foundation of China (Grant Nos.U2167217,12205286,and 11905025)the National MCF Energy Research and Development Program of China (Grant No.2018YFE0308105)。
文摘Numerous irradiation-induced gas bubbles are created in the nuclear fuel during irradiation, leading to the change of microstructure and the degradation of mechanical and thermal properties. The grain size of fuel is one of the important factors affecting bubble evolution. In current study, we first predict the thermodynamic behaviors of point defects as well as the interplay between vacancy and gas atom in both UO_(2) and U_(3)Si_(2) according to ab initio approach. Then, we establish the irradiation-induced bubble phase-field model to investigate the formation and evolution of intra-and inter-granular gas bubbles. The effects of fission rate and temperature on the evolutions of bubble morphologies in UO_(2) and U_(3)Si_(2) have been revealed. Especially, a comparison of porosities under different grain sizes is examined and analyzed. To understand the thermal conductivity as functions of grain size and porosity, the heat transfer capability of U_(3)Si_(2) is evaluated.
文摘This paper presents the ageing mechanism of fuse in nuclear power plant in detail. Metal Electromigration is identified as the dominant ageing mechanism. On this basis, the dominant status indicators, temperature and resistance of fuse were ensured, and current-temperature curve was proposed. The infrared thermal imaging technology was used to inspect the ageing condition and prove the current-temperature curve. Finally, the accelerated ageing testing was conducted abiding by the dominant ageing mechanism, and the lifetime was evaluated.
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 field of nuclear agriculture is introduced very briefly. Nuclear agriculture research/development system in Pakistan is described highlighting the achievements of the system partners. Description and discussion are generalized in concluding remarks at the end of the article. This article is an experimental guide for a developing nuclear agriculture system.
基金financially supported by the National Natural Science Foundation of China(No.52122103)the Sichuan Science and Technology Program(No.2022YFG0346).
文摘FeCrAl alloy is one of the most promising candidates as an accident-tolerant fuel(ATF)cladding material.Herein,the influence of cold-rolling(CR)reduction on microstructure and tensile properties of the as-annealed FeCrAl alloys,with low Cr and Nb contents,is systematically examined.With the increase in CR reduction,the grain size of FeCrAl alloy is obviously refined after annealing because the increase in stored deformation energy leads to enhanced recrystallization.However,the large CR reductions result in a severe mixed-grain microstructure,significantly reducing the uniform deformability of the FeCrAl alloy.The dislocation density of the as-annealed FeCrAl alloy decreases with the increase in CR reduction,except for the excessive CR reduction of 50%.Moreover,the Laves phases are crushed and dissolved during CR and annealing,as well as large amounts of refined Laves phases are found after large CR reductions.The pinning effect of the Laves phases can significantly improve the strength of FeCrAl alloy.Accordingly,the strengthening mechanisms of FeCrAl alloy consist of fine-grain strengthening,dislocation strengthening and precipitation strengthening.Finally,the FeCrAl alloy,with a CR reduction of 30%,achieves optimal tensile properties.This study can provide theoretical guidance for the industrial production of the FeCrAl alloy.
基金the Independent Deployment Project of Fujian Science&Technology Innovation Laboratory for Optoelectronic Information of China(2021ZZ109)the National Key Research and Development Program of China(2019YFC0605000)+2 种基金the Independent Deployment Project of Ganjiang Innovation Research Institute of Chinese Academy of Sciences(E055A002)the Key Deployment Project of the Chinese Academy of Sciences(ZDRW-CN-2021-3)the Fujian Provincial Natural Fund Project(2021J05101)。
文摘In this work,neodymium(Nd)and samarium(Sm)codoped lead lanthanum zirconate titanate(PLZT)ceramics were prepared by a high-temperature solid-state method.The samples were characterized by X-ray diffraction,scanning electron microscopy and ferroelectric analysis.Rare earth-doped PLZT ceramics show good phase formation.An appropriate rare earth element doping amount increases the densities of PLZT ceramics and reduces their resistivities,which is due to the role of rare earth elements in grain refinement.However,the increase in the amount of grain boundaries caused by grain refinement also affects domain inversion.Therefore,with increasing doping concentration,the remnant polarization of PLZT gradually decreases,and the doping of rare earth elements also slightly reduces the band gap of PLZT.Under irradiation with an X-ray simulated beta source with a particle energy of 10 keV(between the average energies of the beta particles of^3H and^(63)Ni),the ceramic sheets in this work produce current densities of up to 1.38 nA/cm^2.This indicates that Nd and Sm codoped PLZT ceramics have a certain potential for application in betavoltaic batteries.
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.
基金the financial support from the China National Natural Science Foundation project(11675126)Project of China Nuclear Power Innovation Center,China National Nuclear Corporation Science fund for talented young scholars(FY18000120)the National Science and Technology Major Project of the Ministry of Science and Technology of China(2019ZX06004009).
文摘Computational simulation is an important technical means in research of nuclear fuel materials.Since nuclear fuel issues are inherently multi-scopic,it is imperative to study them with multi-scale simulation scheme.At present,the development of multi-scale simulation for nuclear fuel materials calls for a more systematic approach,in which lies the main purpose of this article.The most important thing in multi-scale simulation is to accurately formulate the goals to be achieved and the types of methods to be used.In this regard,we first summarize the basic principles and applicability of the simulation methods which are commonly used in nuclear fuel research and are based on different scales ranging from micro to macro,i.e.First-Principles(FP),Molecular Dynamics(MD),Kinetic Monte Carlo(KMC),Phase Field(PF),Rate Theory(RT),and Finite Element Method(FEM).And then we discuss the major material issues in this field,also ranging from micro-scale to macro-scale and covering both pellets and claddings,with emphasis on what simulation method would be most suitable for solving each of the issues.Finally,we give our prospective analysis and understanding about the feasible ways of multi-scale integration and relevant handicaps and challenges.
基金The work was financially supported by the National Natural Science Foundation of China(Nos.51874023,U1860206)the Fundamental Research Funds for the Central Universities(No.FRF-TP-20-02B)the Recruitment Program of Global Experts.
文摘The high burn-up structure(HBS)is characterized by the grain size of 100-300 nm and a porosity of up to 20%,which is formed at the rim of the nuclear fuel pellet due to 2-3 times higher local burn-up during the in-pile irradiation.HBS is considered a new potential structure for high-performance fuels.However,it is difficult to prepare HBS by conventional sintering methods.In this study,flash sintering was used to prepare HBS using CeO_(2)as a surrogate for a preliminary investigation.A new experimental configuration for rapid sintering of CeO_(2)pellets was provided,in which the green body can be rapidly preheated and pressure-assisted by the induction heating electrodes.An insulated quartz tube was used as the die for the flash sintered samples,allowing the current to flow through the sample and providing a stable condition for applying an external pressure of approximately 5.3-7.0 MPa during flash sintering process.Using an initial electric field of 141 V cm-1 and holding for 1-7 min at the maximum current density of~98 mA mm^(-2),CeO_(2)ceramics with a grain size of 114-282 nm and a relative density of 75.4-99.7%were prepared.The densification and microstructure evolution behaviors during flash sintering in this new experimental configuration have been discussed in detail.This new experimental configuration may provide a promising approach for preparing UO_(2)ceramics and their HBS.
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>
