The temporal and spatial growth behaviour of protein crystals, subject to different cooling strategies in protein crystallisation was investigated. Although the impact of temperature and cooling rate on crystal growth...The temporal and spatial growth behaviour of protein crystals, subject to different cooling strategies in protein crystallisation was investigated. Although the impact of temperature and cooling rate on crystal growth of small molecules was well documented, much less has been reported on their impact on the crystallisation of proteins. In this paper, an experimental set-up is configured to carry out such a study which involves an automatic temperature controlled hot-stage crystalliser fitted with a real-time imaging system. Linbro parallel crystallisation experiments(24-well plate) were also conducted to find the suitable initial conditions to be used in the hot-stage crystallisation experiments, including the initial concentration of HEW lysozyme solutions, precipitate concentration and pH value. It was observed that fast cooling rates at the early stage led to precipitates while slow cooling rates produced crystal nuclei, and very slow cooling rates, much smaller than for small molecules are critical to the growth of the nuclei and the crystals to a desired shape. The interesting results provide valuable insight as well as experimental proof of the feasibility and effectiveness of cooling as a means for achieving controlled protein crystallisation, compared with the evaporation approach which was widely used to grow single large crystals for X-ray diffraction study. Since cooling rate control can be easily achieved and has good repeatability, it suggests that large-scale production of protein crystals can be effectively achieved by manipulating cooling rates.展开更多
Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. ...Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. A fast reactor system is one of the most promising options for electricity generation with an efficient utilization of uranium resources and a reduction of radioactive wastes. Based on the experiences gained during the development of the conceptual designs for KALIMER （Korea advanced liquid metal reactor）, the KAERI （Korea Atomic Energy Research Institute） is currently developing advanced SFR （sodium cooled fast reactor） design concepts that can better meet the Gen IV （Generation IV） technology goals. The long-term advanced SFR development plan will be carried out toward the construction of an advanced SFR demonstration plant by 2028. Advanced concept design studies and the development of the advanced SFR technologies necessary for its commercialization and basic key technologies carried out by KAERI are included in this paper.展开更多
Various sources of solid particles might exist in the coolant flow of a liquid metal cooled fast reactor(e.g.,through chemical interaction between the coolant and impurities,air,or water,through corrosion of structura...Various sources of solid particles might exist in the coolant flow of a liquid metal cooled fast reactor(e.g.,through chemical interaction between the coolant and impurities,air,or water,through corrosion of structural materials,or from damaged/molten fuel).Such particles may cause flow blockage accidents in a fuel assembly,resulting in a reduction in coolant flow,which potentially causes a local temperature rise in the fuel cladding,cladding failure,and fuel melt.To understand the blockage formation mechanism,in this study,a series of simulated experiments was conducted by releasing different solid particles from a release device into a reducer pipe using gravity.Through detailed analyses,the influence of various experimental parameters(e.g.,particle diameter,capacity,shape,and static friction coefficient,and the diameter and height of the particle release nozzle)on the blockage characteristics(i.e.,blockage probability and position)was examined.Under the current range of experimental conditions,the blockage was significantly influenced by the aforementioned parameters.The ratio between the particle diameter and outlet size of the reducer pipe might be one of the determining factors governing the occurrence of blockage.Specifically,increasing the ratio enhanced blockage(i.e.,larger probability and higher position within the reducer pipe).Increasing the particle size,particle capacity,particle static friction coefficient,and particle release nozzle diameter led to a rise in the blockage probability;however,increasing the particle release nozzle height had a downward influence on the blockage probability.Finally,blockage was more likely to occur in non-spherical particles case than that of spherical particles.This study provides a large experimental database to promote an understanding of the flow blockage mechanism and improve the validation process of fast reactor safety analysis codes.展开更多
Supercritical water reactor(SCWR)was proposed as a GenerationⅣconcept for building large capacity nuclear power plants.Comparing with the present GenerationⅡandⅢlight water reactors,SCWR possesses great advantages ...Supercritical water reactor(SCWR)was proposed as a GenerationⅣconcept for building large capacity nuclear power plants.Comparing with the present GenerationⅡandⅢlight water reactors,SCWR possesses great advantages of 10%higher efficiency,simpler system design,better sustainability,and so on.However,the selection of materials for fuel cladding and reactor internals of SCWR is facing a great challenge.Corrosion in supercritical steam is of the first important issue to be solved to meet the stringent requirement of the reactor internal components.Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor(SCWR)in static and re-circulating autoclave at the temperatures of 550,600 and 650℃,pressure of about 25 MPa,deaerated or saturated dissolved hydrogen(STP).Nickel base alloy type Hastelloy C276,austenitic stainless steels type 304NG,AL-6XN,HR3C.NF709 and SAVE 25,ferritic/martensitic(F/M)steel type P92,P122 and 410,and oxide dispersion strengthened steel MA 956,are tested.This paper presents corrosion rate,and focuses on the formation and breakdown of corrosion oxide film,and proposes the future trend for the development of SCWR internal structure materials.展开更多
The success of high temperature gas cooled reactor depends upon the safety and quality of the coated particle fuel. The understanding and evaluation of this fuel requires the development of an integrated mechanistic f...The success of high temperature gas cooled reactor depends upon the safety and quality of the coated particle fuel. The understanding and evaluation of this fuel requires the development of an integrated mechanistic fuel performance model that fully describes the mechanical and physicochemical behavior of the fuel particle under irradiation. In this paper, a review of the analytical capability of some of the existing computer codes for coated particle fuel was performed. These existing models and codes include FZJ model, JAERI model, Stress3 model, ATLAS model, PARFUME model and TIMCOAT model. The theoretic model, methodology, calculation parameters and benchmark of these codes were classified. Based on the failure mechanism of coated particle, the advantage and limits of the models were compared and discussed. The calculated results of the coated particles for China HTR-10 by using some existing code are shown. Finally, problems and challenges in fuel performance modeling were listed.展开更多
A new marine propulsion system is proposed . A small liquid sodium cooled reactor acts as prime mover; alkali-metal thermal-to-electric conversion ( AMTEC) cells are employed to convert the heat energy to electricity;...A new marine propulsion system is proposed . A small liquid sodium cooled reactor acts as prime mover; alkali-metal thermal-to-electric conversion ( AMTEC) cells are employed to convert the heat energy to electricity; superconducting magneto-hydrodynamic thruster combined with spray-water thruster works as pr opulsion. The configuration and characteristics of this system are described. Such a nuclear-powered propulsion system is not only free of noise, but also has high reliability and efficiency. It would be a preferable propulsion system for ships in the future.展开更多
Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanimn （HPGe） gamma spectrometer. By using KORIGEN software and MCNP Mon...Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanimn （HPGe） gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different, irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the ~arCs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% （l（r）. The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burimp in future modular pebble bed reactors.展开更多
Differing from traditional pressurized water reactors(PWRs),heat pipe cooled reactors have the unique characteristics of fuel thermal expansion,expansion reactivity feedback,and thermal contact conductance.These react...Differing from traditional pressurized water reactors(PWRs),heat pipe cooled reactors have the unique characteristics of fuel thermal expansion,expansion reactivity feedback,and thermal contact conductance.These reactors require a new multiphysics coupling method.In this paper,a transient coupling method based on OpenFOAM is proposed.The method considers power variation,thermal expansion,heat pipe operation,thermal contact conductance,and gap conductance.In particular,the reactivity feedback caused by working medium redistribution in a heat pipe is also preliminarily considered.A typical heat pipe cooled reactor KRUSTY(Kilowatt Reactor Using Stirling TechnologY)is chosen as the research object.Compared with experimental results of load following,the calculated results are in good agreement and show the validity of the proposed method.To discuss the self-adjusting capability of this type of reactor system,a hypothetical accident is simulated.It is assumed that at the beginning of this accident,loss of the heat sink occurs.After 1500 s of the transient process,the reactor system recovers immediately.During this hypothetical accident,the control rod is always out of the reactor core,and the reactor only relies on the reactivity feedback to regulate the fission power.According to the simulation,the peak temperature is only about 1112 K,which is far below the safety limit.As for system recovery,the reactor needs approximately 2500 s to return to a steady state and can realize effective power regulation by reactivity feedback.This study confirms the availability of this coupling method and that it can be an effective tool for the simulation of heat pipe cooled reactors.展开更多
Much more nuclear energy capacity is needed than currently installed to meet the demand of energy and the requirement on environment protection in the next decades. More stringent nuclear safety standards have to be e...Much more nuclear energy capacity is needed than currently installed to meet the demand of energy and the requirement on environment protection in the next decades. More stringent nuclear safety standards have to be established for future nuclear power plants.The philosophy of a catastrophe free nuclear technology is presented in this paper. The issue of afterheat removal of high temperature gas cooled reactors is handled.It is a striking inherent safety feature of the modular high temperature gas cooled reactor design that the afterheat removal takes place without any active core cooling systems.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
A 10MW High Temperature Gas Cooled Reactor (HTR-10) designed by the Institute of Nuclear Energy Technology (INET) is now being constructed. The steam generator (SG) in the HTR-10 is one of the most important component...A 10MW High Temperature Gas Cooled Reactor (HTR-10) designed by the Institute of Nuclear Energy Technology (INET) is now being constructed. The steam generator (SG) in the HTR-10 is one of the most important components for reactor safety. The thermal-hydraulic performance of the SG was investigated. A full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 thermal hydraulic experimental system in detail. The SGTM-10 simulates the actual thermal and structural parameters of the HTR-10. The SGTM-10 includes three separated loops: the primary helium loop, the secondary water loop, and the tertiary cooling water loop. Two parallel tubes are arranged in the test assembly. The main experimental equipment is shown in the paper. Expermental results are given illustrating the effects of the outlet pressures, the heating power, and the inlet subcooling.展开更多
基金Supported by the China One Thousand Talent Scheme,the National Natural Science Foundation of China under its Major Research Scheme of Meso-scale Mechanism and Control in Multi-phase Reaction Processes(91434126)the Natural Science Foundation of Guangdong Province(2014A030313228)+1 种基金benefited from early work funded by UK Engineering and Physical Science Research Council(EP/H008012/1EP/H008853/1)
文摘The temporal and spatial growth behaviour of protein crystals, subject to different cooling strategies in protein crystallisation was investigated. Although the impact of temperature and cooling rate on crystal growth of small molecules was well documented, much less has been reported on their impact on the crystallisation of proteins. In this paper, an experimental set-up is configured to carry out such a study which involves an automatic temperature controlled hot-stage crystalliser fitted with a real-time imaging system. Linbro parallel crystallisation experiments(24-well plate) were also conducted to find the suitable initial conditions to be used in the hot-stage crystallisation experiments, including the initial concentration of HEW lysozyme solutions, precipitate concentration and pH value. It was observed that fast cooling rates at the early stage led to precipitates while slow cooling rates produced crystal nuclei, and very slow cooling rates, much smaller than for small molecules are critical to the growth of the nuclei and the crystals to a desired shape. The interesting results provide valuable insight as well as experimental proof of the feasibility and effectiveness of cooling as a means for achieving controlled protein crystallisation, compared with the evaporation approach which was widely used to grow single large crystals for X-ray diffraction study. Since cooling rate control can be easily achieved and has good repeatability, it suggests that large-scale production of protein crystals can be effectively achieved by manipulating cooling rates.
文摘Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. A fast reactor system is one of the most promising options for electricity generation with an efficient utilization of uranium resources and a reduction of radioactive wastes. Based on the experiences gained during the development of the conceptual designs for KALIMER （Korea advanced liquid metal reactor）, the KAERI （Korea Atomic Energy Research Institute） is currently developing advanced SFR （sodium cooled fast reactor） design concepts that can better meet the Gen IV （Generation IV） technology goals. The long-term advanced SFR development plan will be carried out toward the construction of an advanced SFR demonstration plant by 2028. Advanced concept design studies and the development of the advanced SFR technologies necessary for its commercialization and basic key technologies carried out by KAERI are included in this paper.
基金supported by the Basic and Applied Basic Research Foundation of Guangdong Province(Nos.2021A1515010343,2022A1515011582)the Science and Technology Program of Guangdong Province(Nos.2021A0505030026,2022A0505050029).
文摘Various sources of solid particles might exist in the coolant flow of a liquid metal cooled fast reactor(e.g.,through chemical interaction between the coolant and impurities,air,or water,through corrosion of structural materials,or from damaged/molten fuel).Such particles may cause flow blockage accidents in a fuel assembly,resulting in a reduction in coolant flow,which potentially causes a local temperature rise in the fuel cladding,cladding failure,and fuel melt.To understand the blockage formation mechanism,in this study,a series of simulated experiments was conducted by releasing different solid particles from a release device into a reducer pipe using gravity.Through detailed analyses,the influence of various experimental parameters(e.g.,particle diameter,capacity,shape,and static friction coefficient,and the diameter and height of the particle release nozzle)on the blockage characteristics(i.e.,blockage probability and position)was examined.Under the current range of experimental conditions,the blockage was significantly influenced by the aforementioned parameters.The ratio between the particle diameter and outlet size of the reducer pipe might be one of the determining factors governing the occurrence of blockage.Specifically,increasing the ratio enhanced blockage(i.e.,larger probability and higher position within the reducer pipe).Increasing the particle size,particle capacity,particle static friction coefficient,and particle release nozzle diameter led to a rise in the blockage probability;however,increasing the particle release nozzle height had a downward influence on the blockage probability.Finally,blockage was more likely to occur in non-spherical particles case than that of spherical particles.This study provides a large experimental database to promote an understanding of the flow blockage mechanism and improve the validation process of fast reactor safety analysis codes.
文摘Supercritical water reactor(SCWR)was proposed as a GenerationⅣconcept for building large capacity nuclear power plants.Comparing with the present GenerationⅡandⅢlight water reactors,SCWR possesses great advantages of 10%higher efficiency,simpler system design,better sustainability,and so on.However,the selection of materials for fuel cladding and reactor internals of SCWR is facing a great challenge.Corrosion in supercritical steam is of the first important issue to be solved to meet the stringent requirement of the reactor internal components.Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor(SCWR)in static and re-circulating autoclave at the temperatures of 550,600 and 650℃,pressure of about 25 MPa,deaerated or saturated dissolved hydrogen(STP).Nickel base alloy type Hastelloy C276,austenitic stainless steels type 304NG,AL-6XN,HR3C.NF709 and SAVE 25,ferritic/martensitic(F/M)steel type P92,P122 and 410,and oxide dispersion strengthened steel MA 956,are tested.This paper presents corrosion rate,and focuses on the formation and breakdown of corrosion oxide film,and proposes the future trend for the development of SCWR internal structure materials.
文摘The success of high temperature gas cooled reactor depends upon the safety and quality of the coated particle fuel. The understanding and evaluation of this fuel requires the development of an integrated mechanistic fuel performance model that fully describes the mechanical and physicochemical behavior of the fuel particle under irradiation. In this paper, a review of the analytical capability of some of the existing computer codes for coated particle fuel was performed. These existing models and codes include FZJ model, JAERI model, Stress3 model, ATLAS model, PARFUME model and TIMCOAT model. The theoretic model, methodology, calculation parameters and benchmark of these codes were classified. Based on the failure mechanism of coated particle, the advantage and limits of the models were compared and discussed. The calculated results of the coated particles for China HTR-10 by using some existing code are shown. Finally, problems and challenges in fuel performance modeling were listed.
文摘A new marine propulsion system is proposed . A small liquid sodium cooled reactor acts as prime mover; alkali-metal thermal-to-electric conversion ( AMTEC) cells are employed to convert the heat energy to electricity; superconducting magneto-hydrodynamic thruster combined with spray-water thruster works as pr opulsion. The configuration and characteristics of this system are described. Such a nuclear-powered propulsion system is not only free of noise, but also has high reliability and efficiency. It would be a preferable propulsion system for ships in the future.
基金Supported by National Science and Technology Major Project(ZX06901)National Natural Science Foundation of China(10975083,11079025)Tsinghua University Initiative Scientific Research Program
文摘Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanimn （HPGe） gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different, irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the ~arCs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% （l（r）. The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burimp in future modular pebble bed reactors.
基金supported by the National Key Research and Development Project of China(Grant No.2020YFB1901700)Science Challenge Project(Grant No.TZ2018001)+1 种基金the National Natural Science Foundation of China(Grant Nos.11775126 and 11775127)the Tsinghua University Initiative Scientific Research Program。
文摘Differing from traditional pressurized water reactors(PWRs),heat pipe cooled reactors have the unique characteristics of fuel thermal expansion,expansion reactivity feedback,and thermal contact conductance.These reactors require a new multiphysics coupling method.In this paper,a transient coupling method based on OpenFOAM is proposed.The method considers power variation,thermal expansion,heat pipe operation,thermal contact conductance,and gap conductance.In particular,the reactivity feedback caused by working medium redistribution in a heat pipe is also preliminarily considered.A typical heat pipe cooled reactor KRUSTY(Kilowatt Reactor Using Stirling TechnologY)is chosen as the research object.Compared with experimental results of load following,the calculated results are in good agreement and show the validity of the proposed method.To discuss the self-adjusting capability of this type of reactor system,a hypothetical accident is simulated.It is assumed that at the beginning of this accident,loss of the heat sink occurs.After 1500 s of the transient process,the reactor system recovers immediately.During this hypothetical accident,the control rod is always out of the reactor core,and the reactor only relies on the reactivity feedback to regulate the fission power.According to the simulation,the peak temperature is only about 1112 K,which is far below the safety limit.As for system recovery,the reactor needs approximately 2500 s to return to a steady state and can realize effective power regulation by reactivity feedback.This study confirms the availability of this coupling method and that it can be an effective tool for the simulation of heat pipe cooled reactors.
文摘Much more nuclear energy capacity is needed than currently installed to meet the demand of energy and the requirement on environment protection in the next decades. More stringent nuclear safety standards have to be established for future nuclear power plants.The philosophy of a catastrophe free nuclear technology is presented in this paper. The issue of afterheat removal of high temperature gas cooled reactors is handled.It is a striking inherent safety feature of the modular high temperature gas cooled reactor design that the afterheat removal takes place without any active core cooling systems.
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the State High- Tech Developments Plan of China!(No.86 3- 6 14- 0 2
文摘A 10MW High Temperature Gas Cooled Reactor (HTR-10) designed by the Institute of Nuclear Energy Technology (INET) is now being constructed. The steam generator (SG) in the HTR-10 is one of the most important components for reactor safety. The thermal-hydraulic performance of the SG was investigated. A full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 thermal hydraulic experimental system in detail. The SGTM-10 simulates the actual thermal and structural parameters of the HTR-10. The SGTM-10 includes three separated loops: the primary helium loop, the secondary water loop, and the tertiary cooling water loop. Two parallel tubes are arranged in the test assembly. The main experimental equipment is shown in the paper. Expermental results are given illustrating the effects of the outlet pressures, the heating power, and the inlet subcooling.
