Nuclear power plants exhibit non-linear and time-variable dynamics.Therefore,designing a control system that sets the reactor power and forces it to follow the desired load is complicated.A supercritical water reactor...Nuclear power plants exhibit non-linear and time-variable dynamics.Therefore,designing a control system that sets the reactor power and forces it to follow the desired load is complicated.A supercritical water reactor(SCWR)is a fourth-generation conceptual reactor.In an SCWR,the non-linear dynamics of the reactor require a controller capable of control-ling the nonlinearities.In this study,a pressure-tube-type SCWR was controlled during reactor power maneuvering with a higher order sliding mode,and the reactor outgoing steam temperature and pressure were controlled simultaneously.In an SCWR,the temperature,pressure,and power must be maintained at a setpoint(desired value)during power maneuvering.Reactor point kinetics equations with three groups of delayed neutrons were used in the simulation.Higher-order and classic sliding mode controllers were separately manufactured to control the plant and were compared with the PI controllers speci-fied in previous studies.The controlled parameters were reactor power,steam temperature,and pressure.Notably,for these parameters,the PI controller had certain instabilities in the presence of disturbances.The classic sliding mode controller had a higher accuracy and stability;however its main drawback was the chattering phenomenon.HOSMC was highly accurate and stable and had a small computational cost.In reality,it followed the desired values without oscillations and chattering.展开更多
This work presents results on the commissioning of the new heat exchanger of the IEA-R1 nuclear reactor in the occasion of its operational power upgrade from 2 MW to 5 MW, in comparison to the values calculated in the...This work presents results on the commissioning of the new heat exchanger of the IEA-R1 nuclear reactor in the occasion of its operational power upgrade from 2 MW to 5 MW, in comparison to the values calculated in the project of IESA Design and Equipments Company. This reactor is a swimming pool type, light water moderated and with graphite reflectors, used for research purposes and medical radioisotopes production. During monitoring procedures, issues were observed on the reactor operation at 5 MW mainly due to the ageing of the reactor＇s oldest heat exchanger （TC-A） and excessive vibrations at high flow rates on the other installed heat exchanger （TC-B）. So it was decided to provide a new IESA heat exchanger with 5 MW capacity to definitely substitute the TC-A heat exchanger. The results show that the IEA-R1 nuclear reactor can be operated safely and continuously at 5 MW with the new IESA heat exchanger.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 mechanical and fatigue properties of SA508-Ⅳ steel with martensite and granular bainite, respectively, were studied. The mechanical tests results showed that the ultimate tensile strength and impact toughness of ...The mechanical and fatigue properties of SA508-Ⅳ steel with martensite and granular bainite, respectively, were studied. The mechanical tests results showed that the ultimate tensile strength and impact toughness of the specimen with martensite were 830 MPa and 158 J, respectively, and those of the specimen with granular bainite were 811 MPa and 115 J, respectively. The former had higher tensile strength and impact toughness than the latter. The impact tests results showed that the former belonged to typical dimple fracture, while the latter belonged to brittle fracture. The fatigue tests results showed that the fatigue life of the specimen with martensite was 2717 cycles, and that of the specimen with granular bainite was 1545 cycles under the strain amplitude of ± 0.45%. The specimen with martensite had fewer crack initiation points, narrower fatigue striations separation, and larger volume fraction of high-angle grain boundaries than the latter. The fewer crack initiation points meant fewer fatigue cracks, the narrower fatigue striations separation meant slower crack propagation rate, and the larger volume fraction of high-angle grain boundaries could more effectively hinder fatigue crack propagation. Based on these facts, the fatigue life of the specimen with martensite was higher than that of the specimen with granular bainite.展开更多
An AP1000 nuclear reactor coolant pump is considered to assess the influence of the Impeller/Guide vane clearance on the performances of this type of pumps.Experiments and numerical simulations relying on an unidirect...An AP1000 nuclear reactor coolant pump is considered to assess the influence of the Impeller/Guide vane clearance on the performances of this type of pumps.Experiments and numerical simulations relying on an unidirectional fluid-solid coupling approach are used to investigate the problem(stress,strain and mode of the rotor).The results reveal the relationship existing between the hydraulic performance of the nuclear reactor coolant pump and the clearance ratio.The effect of clearance ratio on the maximum equivalent stress on the back surface of the impeller blade is greater than that on the working surface(the maximum equivalent stress on the back surface of impeller blade is about three times that on the working surface).The clearance ratio has a scarce effect on the first six natural frequencies of the rotor of the nuclear reactor coolant pump.The related vibrational modes have different waveforms.展开更多
Gas irradiation in research nuclear reactors is an important way to produce radionuclides. Although some nuclear reactors centers offer this type of service, there are few publications about capsules to irradiation of...Gas irradiation in research nuclear reactors is an important way to produce radionuclides. Although some nuclear reactors centers offer this type of service, there are few publications about capsules to irradiation of gaseous samples. This paper describes a method to fabricate and evaluate aluminum capsules to irradiate gaseous samples in nuclear reactor. A semi-circular slotted die from a hydraulic presshead was modified to seal aluminum tubes. The aluminum capsules were subjected to leak detection tests, which demonstrated the accordance with standard ISO 9978.展开更多
With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), w...With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), with the implementation of sodium–potassium eutectic alloy(NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal–hydraulic characteristics of the space nuclear reactor TOPAZ-Ⅱ.A RELAP5 model including thermionic fuel elements(TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector,moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-Ⅱ system model, an electrical ground test of the fully integrated TOPAZ-Ⅱ system, the V-71 unit,is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal–hydraulic characteristics of TOPAZ-Ⅱ are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.展开更多
Advanced nuclear reactors offer safe, clean, and reliable energy at the global scale. The development of such devices relies heavily upon computational models, from the pre-conceptual stages through detailed design, l...Advanced nuclear reactors offer safe, clean, and reliable energy at the global scale. The development of such devices relies heavily upon computational models, from the pre-conceptual stages through detailed design, licensing, and operation. An integrated reactor modeling framework that enables seamless communication, coupling, automation, and continuous development brings significant new capabilities and efficiencies to the practice of reactor design. In such a system, key performance metrics （e.g., optimal fuel management, peak cladding temperature in design-basis accidents, levelized cost of electricity） can be explicitly linked to design inputs （e.g., assembly duct thickness, tolerances）, enabling an exceptional level of design consistency. Coupled with high-performance computing, thousands of integrated cases can be executed simultaneously to analyze the full system, perform complete sensitivity studies, and efficiently and robustly evaluate various design tradeoffs. TerraPower has developed such a tool-the Advanced Reactor Modeling Interface （ARMI） code system-and has deployed it to support the TerraPower Traveling Wave Reactor design and other innovative energy products currently under development. The ARMI code system employs pre-existing tools with strong pedigrees alongside many new physics and data management modules necessary for innovative design. Verification and validation against previous and new physical measurements, which remain an essential element of any sound design, are being carried out. This paper summarizes the integrated core engineering tools and practices in production at TerraPower.展开更多
In nuclear reactors cooled by liquid metals, ultrasound is the only type of field that allows obtaining images of the reactor cores and diagnostics of the integrity of the fuel assemblies. The article discusses the fe...In nuclear reactors cooled by liquid metals, ultrasound is the only type of field that allows obtaining images of the reactor cores and diagnostics of the integrity of the fuel assemblies. The article discusses the features of the practical realization of ultrasonic imaging systems based on phased arrays and offers an alternative solution of imaging on the basis of the acoustic lenses of refractive and diffraction types. Using lenses eliminates many of the technical and technological problems associated with the development of multi-element phased arrays. It is shown that lens systems allow using traditional methods of transformation of acoustic fields into the visible images by 2D piezo matrix and a more promising way of acoustooptical transformation based on coherent optical interferometry.展开更多
The studies performed in the frame of a project destined for the search of new (t,n) and (p,n) reactions of interest in nuclear reactors are described. Experimental evidences of the observations of the reactions: 46Ti...The studies performed in the frame of a project destined for the search of new (t,n) and (p,n) reactions of interest in nuclear reactors are described. Experimental evidences of the observations of the reactions: 46Ti(t,n)48V, 48Ti(p,n)48V, 52Cr(t,n)54Mn, 56Fe(p,n)56Co, 72Ge(t,n)74As and 74Ge(p,n)74As, are presented. Additional data on some secondary reactions, already characterised for the production of 7Be, 56Co, 58Co, 65Zn and 88Y, were also obtained. The significance of these data is discussed.展开更多
Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. They are factory built, requiring smaller initial capital investment and facilitating shorter co...Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. They are factory built, requiring smaller initial capital investment and facilitating shorter construction times. SMRs also promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are public parameters for many near-term SMR projects. The fuel cost of electricity generation for selected SMRs and large reactors is simulated, including calculation of optimal tails assay in the uranium enrichment process. The results are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 70% higher fuel costs than large light water reactors using 2014 nuclear fuels market data. Fuel cost sensitivities to reactor design parameters are presented.展开更多
Compared to other energy sources,nuclear reactors offer several advantages as a spacecraft power source,including compact size,high power density,and long operating life.These qualities make nuclear power an ideal ene...Compared to other energy sources,nuclear reactors offer several advantages as a spacecraft power source,including compact size,high power density,and long operating life.These qualities make nuclear power an ideal energy source for future deep space exploration.A whole system model of the space nuclear reactor consisting of the reactor neutron kinetics,reactivity control,reactor heat transfer,heat exchanger,and thermoelectric converter was developed.In addition,an electrical power control system was designed based on the developed dynamic model.The GRS method was used to quantitatively calculate the uncertainty of coupling parameters of the neutronics,thermal-hydraulics,and control system for the space reactor.The Spearman correlation coefficient was applied in the sensitivity analysis of system input parameters to output parameters.The calculation results showed that the uncertainty of the output parameters caused by coupling parameters had the most considerable variation,with a relative standard deviation<2.01%.Effective delayed neutron fraction was most sensitive to electrical power.To obtain optimal control performance,the non-dominated sorting genetic algorithm method was employed to optimize the controller parameters based on the uncertainty quantification calculation.Two typical transient simulations were conducted to test the adaptive ability of the optimized controller in the uncertainty dynamic system,including 100%full power(FP)to 90%FP step load reduction transient and 5%FP/min linear variable load transient.The results showed that,considering the influence of system uncertainty,the optimized controller could improve the response speed and load following accuracy of electrical power control,in which the effectiveness and superiority have been verified.展开更多
Core power is a key parameter of nuclear reactor.Traditionally,the proportional-integralderivative(PID)controllers are used to control the core power.Fractional-order PID(FOPID)controller represents the cutting edge i...Core power is a key parameter of nuclear reactor.Traditionally,the proportional-integralderivative(PID)controllers are used to control the core power.Fractional-order PID(FOPID)controller represents the cutting edge in core power control research.In comparing with the integer-order models,fractional-order models describe the variation of core power more accurately,thus provide a comprehensive and realistic depiction for the power and state changes of reactor core.However,current fractional-order controllers cannot adjust their parameters dynamically to response the environmental changes or demands.In this paper,we aim at the stable control and dynamic responsiveness of core power.Based on the strong selflearning ability of artificial neural network(ANN),we propose a composite controller combining the ANN and FOPID controller.The FOPID controller is firstly designed and a back propagation neural network(BPNN)is then utilized to optimize the parameters of FOPID.It is shown by simulation that the composite controller enables the real-time parameter tuning via ANN and retains the advantage of FOPID controller.展开更多
Accurate water level measurement in nuclear reactors,particularly in PWRs(pressurized water reactors)and BWRs(boiling water reactors),is essential for ensuring the safety and efficiency of reactor operations.K-type HJ...Accurate water level measurement in nuclear reactors,particularly in PWRs(pressurized water reactors)and BWRs(boiling water reactors),is essential for ensuring the safety and efficiency of reactor operations.K-type HJTCs(heated junction thermocouples)are widely used for this purpose due to their ability to withstand extreme temperatures and radiation conditions.This article explores the role of HJTCs in reactor water level measurement and compares the performance of 2-wire and 3-wire connections.While the 2-wire connection is simple and cost-effective,it can introduce measurement inaccuracies due to wire resistance.In contrast,the 3-wire connection compensates for lead resistance,offering more precise and reliable measurements,particularly in long-distance applications.This paper discusses the operational considerations of these wiring configurations in the context of nuclear reactors and highlights the importance of choosing the appropriate connection type to optimize safety and measurement accuracy in PWR and BWR reactors.展开更多
Nuclear reactor safety(NRS)and the branch accident analysis(AA)constitute proven technologies:these are based on,among the other things,long lasting research and operational experience in the area of water cooled nucl...Nuclear reactor safety(NRS)and the branch accident analysis(AA)constitute proven technologies:these are based on,among the other things,long lasting research and operational experience in the area of water cooled nuclear reactors(WCNR).Large break loss of coolant accident(LBLOCA)has been,so far,the orienting scenario within AA and a basis for the design of reactors.An incomplete vision for those technologies during the last few years is as follows:Progress in fundamentals was stagnant,namely in those countries where the WCNR were designed.Weaknesses became evident,noticeably in relation to nuclear fuel under high burn-up.Best estimate plus uncertainty(BEPU)techniques were perfected and available for application.Electronic and informatics systems were in extensive use and their impact in case of accident becomes more and more un-checked(however,quite irrelevant in case of LBLOCA).The time delay between technological discoveries and applications was becoming longer.The present paper deals with the LBLOCA that is inserted into the above context.Key conclusion is that regulations need suitable modification,rather than lowering the importance and the role of LBLOCA.Moreover,strengths of emergency core cooling system(ECCS)and containment need a tight link.展开更多
The dual-cooled nuclear reactor is currently considered for improving the designs of current/future nuclear reactors. Investigation of the thermal-hydraulic characteristics of the nuclear reactor via experiments is es...The dual-cooled nuclear reactor is currently considered for improving the designs of current/future nuclear reactors. Investigation of the thermal-hydraulic characteristics of the nuclear reactor via experiments is essential for commercializing the dual-cooled nuclear reactor. In this paper, the turbulent flow in square arrayed six-rod bundles in the form of magnified copies of the dual-cooled and current OPR-1000 nuclear reactor is experimentally investigated by means of hot-wire anemometry and smoke-wire generation methods. Vortex trains which do not exist in an ordinary reactor subchannel are presented in the subchannel of the dual-cooled reactor. The vortices are induced by a span-wise velocity gradient. This flow pulsation phenomenon increases the inter-channel mixing of the subchannel. To understand the periodic feature of the pulsation, axial/cross velocities are measured and the periodic characteristic frequencies are obtained by a Fast Fourier Transform (FFT) analysis. The peak frequency that represents the quasi-periodic pulsation of the flow is increased with an increase in the axial velocity while the wavelength of the pulsation remains constant within a tested range of the Reynolds number (9000 51000). The vortex trains are highly synchronized with each other, as confirmed by means of visualization.展开更多
The aging of operational reactors leads to increased mechanical vibrations in the reactor interior.The vibration of the incore sensors near their nominal locations is a new problem for neutronic field reconstruction.C...The aging of operational reactors leads to increased mechanical vibrations in the reactor interior.The vibration of the incore sensors near their nominal locations is a new problem for neutronic field reconstruction.Current field-reconstruction methods fail to handle spatially moving sensors.In this study,we propose a Voronoi tessellation technique in combination with convolutional neural networks to handle this challenge.Observations from movable in-core sensors were projected onto the same global field structure using Voronoi tessellation,holding the magnitude and location information of the sensors.General convolutional neural networks were used to learn maps from observations to the global field.The proposed method reconstructed multi-physics fields(including fast flux,thermal flux,and power rate)using observations from a single field(such as thermal flux).Numerical tests based on the IAEA benchmark demonstrated the potential of the proposed method in practical engineering applications,particularly within an amplitude of 5 cm around the nominal locations,which led to average relative errors below 5% and 10% in the L_(2) and L_(∞)norms,respectively.展开更多
The large fire door is the key component to ensure the effectiveness of fire zone in AP1000 nuclear reactor.According to the fire design requirements and design criteria,the global structure of the large fire door is ...The large fire door is the key component to ensure the effectiveness of fire zone in AP1000 nuclear reactor.According to the fire design requirements and design criteria,the global structure of the large fire door is designed.Based on the designed structure,the thermal mathematical model of the large fire door is established.Based on the solid heat transfer theory,the multi-layer heat transfer theory and integrated heat transfer theory,the differential equations of heat conduction,initial conditions,and boundary conditions are determined.Thermal analysis for the fire door leaf and the closure is carried out by using the method of numerical simulation.Results show that:considering the thermal load,the whole structure of the large fire door can meet the fire resistance limit of 3 hours and the design is reasonable and feasible.This study provides theory basis for the design of the large fire door.展开更多
The 21st Century era and new modern technologies surrounding us day-in and day-out have opened a new door to“Pandora Box”,that we do know it as AI(artificial intelligence)and its two essential integrated components ...The 21st Century era and new modern technologies surrounding us day-in and day-out have opened a new door to“Pandora Box”,that we do know it as AI(artificial intelligence)and its two essential integrated components namely ML(machine learning)and DL(deep learning).However,the strive and progress in AI,ML,and DL pretty much has taken over any industry that we can think of,when it comes to dealing with cloud of structured data in form of BD(big data).A NPP(nuclear power plant)has multiple complicated dynamic system-of-components that have nonlinear behaviors.For controlling the plant operation under both normal and abnormal conditions,the different systems in NPPs(e.g.,the reactor core components,primary and secondary coolant systems)are usually monitored continuously,which leads to very huge amounts of data.Of course Nuclear Power Industry in form of GEN-IV(Generation IV)has not been left behind in this 21st century era by moving out of GEN-III(Generation III)to more modulars form of GEN-IV,known as SMRs(small modular reactors),with a lot of electronic gadgets and electronics that read data and information from it to support safety of these reactor,while in operation with a built in PRA(probabilistic risk assessment),which requires augmentation of AI in them to enhance performance of human operators that are engaged with day-to-day smooth operation of these reactors to make them safe and safer as well as resilience against any natural or man-made disasters by obtaining information through ML from DL that is collecting massive stream of data coming via omni-direction.Integration of AI with HI(human intelligence)is not separable,when it comes to operation of these smart SMRs with state of the art and smart control rooms with human in them as actors.This TM(technical memorandum)is describing the necessity of AI playing with nuclear reactor power plant of GEN-IV being in operation within near term sooner than later,when specially we are facing today’s cyber-attacks with their smart malware agents at work.展开更多
文摘Nuclear power plants exhibit non-linear and time-variable dynamics.Therefore,designing a control system that sets the reactor power and forces it to follow the desired load is complicated.A supercritical water reactor(SCWR)is a fourth-generation conceptual reactor.In an SCWR,the non-linear dynamics of the reactor require a controller capable of control-ling the nonlinearities.In this study,a pressure-tube-type SCWR was controlled during reactor power maneuvering with a higher order sliding mode,and the reactor outgoing steam temperature and pressure were controlled simultaneously.In an SCWR,the temperature,pressure,and power must be maintained at a setpoint(desired value)during power maneuvering.Reactor point kinetics equations with three groups of delayed neutrons were used in the simulation.Higher-order and classic sliding mode controllers were separately manufactured to control the plant and were compared with the PI controllers speci-fied in previous studies.The controlled parameters were reactor power,steam temperature,and pressure.Notably,for these parameters,the PI controller had certain instabilities in the presence of disturbances.The classic sliding mode controller had a higher accuracy and stability;however its main drawback was the chattering phenomenon.HOSMC was highly accurate and stable and had a small computational cost.In reality,it followed the desired values without oscillations and chattering.
文摘This work presents results on the commissioning of the new heat exchanger of the IEA-R1 nuclear reactor in the occasion of its operational power upgrade from 2 MW to 5 MW, in comparison to the values calculated in the project of IESA Design and Equipments Company. This reactor is a swimming pool type, light water moderated and with graphite reflectors, used for research purposes and medical radioisotopes production. During monitoring procedures, issues were observed on the reactor operation at 5 MW mainly due to the ageing of the reactor＇s oldest heat exchanger （TC-A） and excessive vibrations at high flow rates on the other installed heat exchanger （TC-B）. So it was decided to provide a new IESA heat exchanger with 5 MW capacity to definitely substitute the TC-A heat exchanger. The results show that the IEA-R1 nuclear reactor can be operated safely and continuously at 5 MW with the new IESA heat exchanger.
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the Beijing Municipal Natural Science Foundation under No.2162026 and the 863 Program of China under Nos.2008AA031702 and 2012AA03A507 for financial support.
文摘The mechanical and fatigue properties of SA508-Ⅳ steel with martensite and granular bainite, respectively, were studied. The mechanical tests results showed that the ultimate tensile strength and impact toughness of the specimen with martensite were 830 MPa and 158 J, respectively, and those of the specimen with granular bainite were 811 MPa and 115 J, respectively. The former had higher tensile strength and impact toughness than the latter. The impact tests results showed that the former belonged to typical dimple fracture, while the latter belonged to brittle fracture. The fatigue tests results showed that the fatigue life of the specimen with martensite was 2717 cycles, and that of the specimen with granular bainite was 1545 cycles under the strain amplitude of ± 0.45%. The specimen with martensite had fewer crack initiation points, narrower fatigue striations separation, and larger volume fraction of high-angle grain boundaries than the latter. The fewer crack initiation points meant fewer fatigue cracks, the narrower fatigue striations separation meant slower crack propagation rate, and the larger volume fraction of high-angle grain boundaries could more effectively hinder fatigue crack propagation. Based on these facts, the fatigue life of the specimen with martensite was higher than that of the specimen with granular bainite.
基金This work is supported by the National Natural Science Foundation of China(No.51469013).
文摘An AP1000 nuclear reactor coolant pump is considered to assess the influence of the Impeller/Guide vane clearance on the performances of this type of pumps.Experiments and numerical simulations relying on an unidirectional fluid-solid coupling approach are used to investigate the problem(stress,strain and mode of the rotor).The results reveal the relationship existing between the hydraulic performance of the nuclear reactor coolant pump and the clearance ratio.The effect of clearance ratio on the maximum equivalent stress on the back surface of the impeller blade is greater than that on the working surface(the maximum equivalent stress on the back surface of impeller blade is about three times that on the working surface).The clearance ratio has a scarce effect on the first six natural frequencies of the rotor of the nuclear reactor coolant pump.The related vibrational modes have different waveforms.
文摘Gas irradiation in research nuclear reactors is an important way to produce radionuclides. Although some nuclear reactors centers offer this type of service, there are few publications about capsules to irradiation of gaseous samples. This paper describes a method to fabricate and evaluate aluminum capsules to irradiate gaseous samples in nuclear reactor. A semi-circular slotted die from a hydraulic presshead was modified to seal aluminum tubes. The aluminum capsules were subjected to leak detection tests, which demonstrated the accordance with standard ISO 9978.
基金supported by the China National Postdoctoral Program for Innovative Talents(No.BX201600124)China Postdoctoral Science Foundation(No.2016M600796)the National Natural Science Foundation of China(No.11605131)
文摘With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), with the implementation of sodium–potassium eutectic alloy(NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal–hydraulic characteristics of the space nuclear reactor TOPAZ-Ⅱ.A RELAP5 model including thermionic fuel elements(TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector,moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-Ⅱ system model, an electrical ground test of the fully integrated TOPAZ-Ⅱ system, the V-71 unit,is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal–hydraulic characteristics of TOPAZ-Ⅱ are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.
文摘Advanced nuclear reactors offer safe, clean, and reliable energy at the global scale. The development of such devices relies heavily upon computational models, from the pre-conceptual stages through detailed design, licensing, and operation. An integrated reactor modeling framework that enables seamless communication, coupling, automation, and continuous development brings significant new capabilities and efficiencies to the practice of reactor design. In such a system, key performance metrics （e.g., optimal fuel management, peak cladding temperature in design-basis accidents, levelized cost of electricity） can be explicitly linked to design inputs （e.g., assembly duct thickness, tolerances）, enabling an exceptional level of design consistency. Coupled with high-performance computing, thousands of integrated cases can be executed simultaneously to analyze the full system, perform complete sensitivity studies, and efficiently and robustly evaluate various design tradeoffs. TerraPower has developed such a tool-the Advanced Reactor Modeling Interface （ARMI） code system-and has deployed it to support the TerraPower Traveling Wave Reactor design and other innovative energy products currently under development. The ARMI code system employs pre-existing tools with strong pedigrees alongside many new physics and data management modules necessary for innovative design. Verification and validation against previous and new physical measurements, which remain an essential element of any sound design, are being carried out. This paper summarizes the integrated core engineering tools and practices in production at TerraPower.
文摘In nuclear reactors cooled by liquid metals, ultrasound is the only type of field that allows obtaining images of the reactor cores and diagnostics of the integrity of the fuel assemblies. The article discusses the features of the practical realization of ultrasonic imaging systems based on phased arrays and offers an alternative solution of imaging on the basis of the acoustic lenses of refractive and diffraction types. Using lenses eliminates many of the technical and technological problems associated with the development of multi-element phased arrays. It is shown that lens systems allow using traditional methods of transformation of acoustic fields into the visible images by 2D piezo matrix and a more promising way of acoustooptical transformation based on coherent optical interferometry.
文摘The studies performed in the frame of a project destined for the search of new (t,n) and (p,n) reactions of interest in nuclear reactors are described. Experimental evidences of the observations of the reactions: 46Ti(t,n)48V, 48Ti(p,n)48V, 52Cr(t,n)54Mn, 56Fe(p,n)56Co, 72Ge(t,n)74As and 74Ge(p,n)74As, are presented. Additional data on some secondary reactions, already characterised for the production of 7Be, 56Co, 58Co, 65Zn and 88Y, were also obtained. The significance of these data is discussed.
文摘Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. They are factory built, requiring smaller initial capital investment and facilitating shorter construction times. SMRs also promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are public parameters for many near-term SMR projects. The fuel cost of electricity generation for selected SMRs and large reactors is simulated, including calculation of optimal tails assay in the uranium enrichment process. The results are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 70% higher fuel costs than large light water reactors using 2014 nuclear fuels market data. Fuel cost sensitivities to reactor design parameters are presented.
基金supported by the National Natural Science Foundation of China(12305185)Natural Science Foundation of Hunan Province,China(No.2023JJ50122)+1 种基金International Cooperative Research Project of the Ministry of Education,China(No.HZKY20220355)Scientific Research Foundation of the Education Department of Hunan Province,China(No.22A0307).
文摘Compared to other energy sources,nuclear reactors offer several advantages as a spacecraft power source,including compact size,high power density,and long operating life.These qualities make nuclear power an ideal energy source for future deep space exploration.A whole system model of the space nuclear reactor consisting of the reactor neutron kinetics,reactivity control,reactor heat transfer,heat exchanger,and thermoelectric converter was developed.In addition,an electrical power control system was designed based on the developed dynamic model.The GRS method was used to quantitatively calculate the uncertainty of coupling parameters of the neutronics,thermal-hydraulics,and control system for the space reactor.The Spearman correlation coefficient was applied in the sensitivity analysis of system input parameters to output parameters.The calculation results showed that the uncertainty of the output parameters caused by coupling parameters had the most considerable variation,with a relative standard deviation<2.01%.Effective delayed neutron fraction was most sensitive to electrical power.To obtain optimal control performance,the non-dominated sorting genetic algorithm method was employed to optimize the controller parameters based on the uncertainty quantification calculation.Two typical transient simulations were conducted to test the adaptive ability of the optimized controller in the uncertainty dynamic system,including 100%full power(FP)to 90%FP step load reduction transient and 5%FP/min linear variable load transient.The results showed that,considering the influence of system uncertainty,the optimized controller could improve the response speed and load following accuracy of electrical power control,in which the effectiveness and superiority have been verified.
文摘Core power is a key parameter of nuclear reactor.Traditionally,the proportional-integralderivative(PID)controllers are used to control the core power.Fractional-order PID(FOPID)controller represents the cutting edge in core power control research.In comparing with the integer-order models,fractional-order models describe the variation of core power more accurately,thus provide a comprehensive and realistic depiction for the power and state changes of reactor core.However,current fractional-order controllers cannot adjust their parameters dynamically to response the environmental changes or demands.In this paper,we aim at the stable control and dynamic responsiveness of core power.Based on the strong selflearning ability of artificial neural network(ANN),we propose a composite controller combining the ANN and FOPID controller.The FOPID controller is firstly designed and a back propagation neural network(BPNN)is then utilized to optimize the parameters of FOPID.It is shown by simulation that the composite controller enables the real-time parameter tuning via ANN and retains the advantage of FOPID controller.
文摘Accurate water level measurement in nuclear reactors,particularly in PWRs(pressurized water reactors)and BWRs(boiling water reactors),is essential for ensuring the safety and efficiency of reactor operations.K-type HJTCs(heated junction thermocouples)are widely used for this purpose due to their ability to withstand extreme temperatures and radiation conditions.This article explores the role of HJTCs in reactor water level measurement and compares the performance of 2-wire and 3-wire connections.While the 2-wire connection is simple and cost-effective,it can introduce measurement inaccuracies due to wire resistance.In contrast,the 3-wire connection compensates for lead resistance,offering more precise and reliable measurements,particularly in long-distance applications.This paper discusses the operational considerations of these wiring configurations in the context of nuclear reactors and highlights the importance of choosing the appropriate connection type to optimize safety and measurement accuracy in PWR and BWR reactors.
基金the Institutional Funds ofUniversity of Pisa,Italy。
文摘Nuclear reactor safety(NRS)and the branch accident analysis(AA)constitute proven technologies:these are based on,among the other things,long lasting research and operational experience in the area of water cooled nuclear reactors(WCNR).Large break loss of coolant accident(LBLOCA)has been,so far,the orienting scenario within AA and a basis for the design of reactors.An incomplete vision for those technologies during the last few years is as follows:Progress in fundamentals was stagnant,namely in those countries where the WCNR were designed.Weaknesses became evident,noticeably in relation to nuclear fuel under high burn-up.Best estimate plus uncertainty(BEPU)techniques were perfected and available for application.Electronic and informatics systems were in extensive use and their impact in case of accident becomes more and more un-checked(however,quite irrelevant in case of LBLOCA).The time delay between technological discoveries and applications was becoming longer.The present paper deals with the LBLOCA that is inserted into the above context.Key conclusion is that regulations need suitable modification,rather than lowering the importance and the role of LBLOCA.Moreover,strengths of emergency core cooling system(ECCS)and containment need a tight link.
基金carried out under the Nuclear R&D Program supported by the Ministry of Education, Science and Technology of the Republic of Korea (Grant No. NRF-2012M2A8A5025824)the National Research Foundation of Korea (NRF) grant funded by the Korea government (MEST) (Grant No. 2012-0005727)
文摘The dual-cooled nuclear reactor is currently considered for improving the designs of current/future nuclear reactors. Investigation of the thermal-hydraulic characteristics of the nuclear reactor via experiments is essential for commercializing the dual-cooled nuclear reactor. In this paper, the turbulent flow in square arrayed six-rod bundles in the form of magnified copies of the dual-cooled and current OPR-1000 nuclear reactor is experimentally investigated by means of hot-wire anemometry and smoke-wire generation methods. Vortex trains which do not exist in an ordinary reactor subchannel are presented in the subchannel of the dual-cooled reactor. The vortices are induced by a span-wise velocity gradient. This flow pulsation phenomenon increases the inter-channel mixing of the subchannel. To understand the periodic feature of the pulsation, axial/cross velocities are measured and the periodic characteristic frequencies are obtained by a Fast Fourier Transform (FFT) analysis. The peak frequency that represents the quasi-periodic pulsation of the flow is increased with an increase in the axial velocity while the wavelength of the pulsation remains constant within a tested range of the Reynolds number (9000 51000). The vortex trains are highly synchronized with each other, as confirmed by means of visualization.
基金partially supported by the Natural Science Foundation of Shanghai(No.23ZR1429300)the Innovation Fund of CNNC(Lingchuang Fund)+1 种基金EP/T000414/1 PREdictive Modeling with QuantIfication of UncERtainty for MultiphasE Systems(PREMIERE)the Leverhulme Centre for Wildfires,Environment,and Society through the Leverhulme Trust(No.RC-2018-023).
文摘The aging of operational reactors leads to increased mechanical vibrations in the reactor interior.The vibration of the incore sensors near their nominal locations is a new problem for neutronic field reconstruction.Current field-reconstruction methods fail to handle spatially moving sensors.In this study,we propose a Voronoi tessellation technique in combination with convolutional neural networks to handle this challenge.Observations from movable in-core sensors were projected onto the same global field structure using Voronoi tessellation,holding the magnitude and location information of the sensors.General convolutional neural networks were used to learn maps from observations to the global field.The proposed method reconstructed multi-physics fields(including fast flux,thermal flux,and power rate)using observations from a single field(such as thermal flux).Numerical tests based on the IAEA benchmark demonstrated the potential of the proposed method in practical engineering applications,particularly within an amplitude of 5 cm around the nominal locations,which led to average relative errors below 5% and 10% in the L_(2) and L_(∞)norms,respectively.
基金the support of the National Natural Science Foundation of China (Grant No. 51672241)Jiangsu Science and Technology Plan Project of China (Grant No. BE2016134)+5 种基金the 14th batch High-level Talents Project for "Six Talents Peak" (Grant No. XCL-092)the Province Postdoctoral Foundation of Jiangsu (1501164B)the Technical Innovation Nurturing Foundation of Yangzhou University (2017CXJ024)China Postdoctoral Science Foundation (2016M600447)Yangzhou Innovative Capacity Building Plan Project (YZ2017275)Yangzhou University Science Foundation Project (x20180290)
文摘The large fire door is the key component to ensure the effectiveness of fire zone in AP1000 nuclear reactor.According to the fire design requirements and design criteria,the global structure of the large fire door is designed.Based on the designed structure,the thermal mathematical model of the large fire door is established.Based on the solid heat transfer theory,the multi-layer heat transfer theory and integrated heat transfer theory,the differential equations of heat conduction,initial conditions,and boundary conditions are determined.Thermal analysis for the fire door leaf and the closure is carried out by using the method of numerical simulation.Results show that:considering the thermal load,the whole structure of the large fire door can meet the fire resistance limit of 3 hours and the design is reasonable and feasible.This study provides theory basis for the design of the large fire door.
文摘The 21st Century era and new modern technologies surrounding us day-in and day-out have opened a new door to“Pandora Box”,that we do know it as AI(artificial intelligence)and its two essential integrated components namely ML(machine learning)and DL(deep learning).However,the strive and progress in AI,ML,and DL pretty much has taken over any industry that we can think of,when it comes to dealing with cloud of structured data in form of BD(big data).A NPP(nuclear power plant)has multiple complicated dynamic system-of-components that have nonlinear behaviors.For controlling the plant operation under both normal and abnormal conditions,the different systems in NPPs(e.g.,the reactor core components,primary and secondary coolant systems)are usually monitored continuously,which leads to very huge amounts of data.Of course Nuclear Power Industry in form of GEN-IV(Generation IV)has not been left behind in this 21st century era by moving out of GEN-III(Generation III)to more modulars form of GEN-IV,known as SMRs(small modular reactors),with a lot of electronic gadgets and electronics that read data and information from it to support safety of these reactor,while in operation with a built in PRA(probabilistic risk assessment),which requires augmentation of AI in them to enhance performance of human operators that are engaged with day-to-day smooth operation of these reactors to make them safe and safer as well as resilience against any natural or man-made disasters by obtaining information through ML from DL that is collecting massive stream of data coming via omni-direction.Integration of AI with HI(human intelligence)is not separable,when it comes to operation of these smart SMRs with state of the art and smart control rooms with human in them as actors.This TM(technical memorandum)is describing the necessity of AI playing with nuclear reactor power plant of GEN-IV being in operation within near term sooner than later,when specially we are facing today’s cyber-attacks with their smart malware agents at work.
