One of the main issues in designing optimum tapered cascades for uranium enrichment for annual fuel production in a power reactor is whether to employ large(fat)or small(thin)cascades.What will be the permissible and ...One of the main issues in designing optimum tapered cascades for uranium enrichment for annual fuel production in a power reactor is whether to employ large(fat)or small(thin)cascades.What will be the permissible and optimal ranges of the number of machines that can be used in a cascade?For the first time,the permissible and optimal ranges of the number of gas centrifuges that can be utilized in a cascade were investigated using two types of centrifuges,and the performance of small and large tapered cascades was discussed.The particle swarm optimization algorithm(PSO)has been used to optimize tapered cascades.The results show:(1)For the first centrifuge,41 cascades(91≤n≤4897)and for the second centrifuge,49 cascades(18≤n≤3839)with small and large sizes can be used in enrichment facilities,and the best cascade for them has 530(with 23 stages)and 39(with 7 stages)centrifuges,respectively.(2)For both centrifuges,when 600≤n(number of centrifuges=n),the large cascade performance changes are relatively insignificant.(3)For both types of gas centrifuges,the annual los s of separation power in enrichment facilities is approximately 1.25%-4.82%of the total separation work required.展开更多
Space nuclear reactor power(SNRP)using a gas-cooled reactor(GCR)and a closed Brayton cycle(CBC)is the ideal choice for future high-power space missions.To investigate the safety characteristics and develop the control...Space nuclear reactor power(SNRP)using a gas-cooled reactor(GCR)and a closed Brayton cycle(CBC)is the ideal choice for future high-power space missions.To investigate the safety characteristics and develop the control strategies for gas-cooled SNRP,transient models for GCR,energy conversion unit,pipes,heat exchangers,pump and heat pipe radiator are established and a system analysis code is developed in this paper.Then,analyses of several operation conditions are performed using this code.In full-power steady-state operation,the core hot spot of 1293 K occurs near the upper part of the core.If 0.4$reactivity is introduced into the core,the maximum temperature that the fuel can reach is 2059 K,which is 914 K lower than the fuel melting point.The system finally has the ability to achieve a new steady-state with a higher reactor power.When the GCR is shut down in an emergency,the residual heat of the reactor can be removed through the conduction of the core and radiation heat transfer.The results indicate that the designed GCR is inherently safe owing to its negative reactivity feedback and passive decay heat removal.This paper may provide valuable references for safety design and analysis of the gas-cooled SNRP coupled with CBC.展开更多
Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and econom...Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and economic operation for nuclear power plants( NPPs) especially for a large number of nuclear powers under construction. Based on the development and application of maintenance template in developed countries,and combining with reliability-centered maintenance( RCM) analysis results and maintenance experience data over the past ten years in domestic NPPs, the development process of maintenance template was presented for Chinese pressurized water reactor( PWR) NPP,and the application of maintenance template to maintenance program development and maintenance optimization combined with cases were demonstrated. A shortcut was provided for improving the efficiency of maintenance optimization in domestic PWR NPP,and help to realize a safe,reliable,and economic operation for domestic NPPs.展开更多
It is very important to reduce the construction duration of the Reactor Containment Building (RCB) when considering the more than 50 months on average from concrete placement to completion. Through a case study, this ...It is very important to reduce the construction duration of the Reactor Containment Building (RCB) when considering the more than 50 months on average from concrete placement to completion. Through a case study, this study performs a pre-study for the reduction of construction duration in nuclear power plant project based on construction process of the RCB. The actual data of the case study have been collected and analyze the process and the external wall drawings of the RCB with construction practitioners. As a result of that, it is necessary to modularize the external wall form for equipment hatch and to extend the height of one layer of the external wall form to reduce the construction duration of RCB. The results of this study will be utilized to reduce construction duration of the nuclear power plant.展开更多
This article looks for the necessary conditions to use Deuterium-Deuterium (D-D) fusion for a large power plant. At the moment, for nearly all the projects (JET, ITER…) only the Deuterium-Tritium (D-T) fuel is consid...This article looks for the necessary conditions to use Deuterium-Deuterium (D-D) fusion for a large power plant. At the moment, for nearly all the projects (JET, ITER…) only the Deuterium-Tritium (D-T) fuel is considered for a power plant. However, as shown in this article, even if a D-D reactor would be necessarily much bigger than a D-T reactor due to the much weaker fusion reactivity of the D-D fusion compared to the D-T fusion, a D-D reactor size would remain under an acceptable size. Indeed, a D-D power plant would be necessarily large and powerful, i.e. the net electric power would be equal to a minimum of 1.2 GWe and preferably above 10 GWe. A D-D reactor would be less complex than a D-T reactor as it is not necessary to obtain Tritium from the reactor itself. It is proposed the same type of reactor yet proposed by the author in a previous article, i.e. a Stellarator “racetrack” magnetic loop. The working of this reactor is continuous. It is reminded that the Deuterium is relatively abundant on the sea water, and so it constitutes an almost inexhaustible source of energy. Thanks to secondary fusions (D-T and D-He3) which both occur at an appreciable level above 100 keV, plasma can stabilize around such high equilibrium energy (i.e. between 100 and 150 keV). The mechanical gain (Q) of such reactor increases with the internal pipe radius, up to 4.5 m. A radius of 4.5 m permits a mechanical gain (Q) of about 17 which thanks to a modern thermo-dynamical conversion would lead to convert about 21% of the thermal power issued from the D-D reactor in a net electric power of 20 GWe. The goal of the article is to create a physical model of the D-D reactor so as to estimate this one without the need of a simulator and finally to estimate the dimensions, power and yield of such D-D reactor for different net electrical powers. The difficulties of the modeling of such reactor are listed in this article and would certainly be applicable to a future D-He3 reactor, if any.展开更多
The paper describes the instrumentation for studying temperature and void reactivity effects that were developed at VR-I zero power reactor. Further are described its operational parameters, fields and ways of its uti...The paper describes the instrumentation for studying temperature and void reactivity effects that were developed at VR-I zero power reactor. Further are described its operational parameters, fields and ways of its utilization as well as issues connected to its implementation into the reactor core.展开更多
This paper introduces the simulation, and controls using Simulink of MATLAB for PCTRAN (Personal Computer Transient Analysis) of the power control system (PWR) type pressurized water reactor of PWR WESTINGHOUSE AP1000...This paper introduces the simulation, and controls using Simulink of MATLAB for PCTRAN (Personal Computer Transient Analysis) of the power control system (PWR) type pressurized water reactor of PWR WESTINGHOUSE AP1000. The power controller model produces mathematical model description in nonlinear relation form in Simulink of MATLAB which is an important and popular program used at most universities for education. The power controller is described by a block diagram in this paper and some details introduce to clearly understand the work function. The results of action control compared with the PCTRAN programme in modes of automatic and manual control.展开更多
HPR1000 is an advanced nuclear power plant(NPP)with the significant feature of an active and passive safety design philosophy,developed by the China National Nuclear Corporation.On one hand,it is an evolutionary desig...HPR1000 is an advanced nuclear power plant(NPP)with the significant feature of an active and passive safety design philosophy,developed by the China National Nuclear Corporation.On one hand,it is an evolutionary design based on proven technology of the existing pressurized water reactor NPP;on the other hand,it incorporates advanced design features including a 177-fuel-assembly core loaded with CF3 fuel assemblies,active and passive safety systems,comprehensive severe accident prevention and mitigation measures,enhanced protection against external events,and improved emergency response capability.Extensive verification experiments and tests have been performed for critical innovative improvements on passive systems,the reactor core,and the main equipment.The design of HPR1000fulfills the international utility requirements for advanced light water reactors and the latest nuclear safety requirements,and addresses the safety issues relevant to the Fukushima accident.Along with its outstanding safety and economy,HPR1000 provides an excellent and practicable solution for both domestic and international nuclear power markets.展开更多
The radioactive isotope 60Co is used in many applications and is typically produced in heavy water reactors.As most of the commercial reactors in operation are pressurized light water reactors(PWRs),the world supply o...The radioactive isotope 60Co is used in many applications and is typically produced in heavy water reactors.As most of the commercial reactors in operation are pressurized light water reactors(PWRs),the world supply of high level radioactive cobalt would be greatly increased if 60Co could be produced in them.Currently,60Co production in PWRs has not been extensively studied;for the 59Co(n,c)60Co reaction,the positioning of 59Co rods in the reactor determines the rate of production.This article primarily uses the models of 60Co production in Canadian CANDU power reactors and American boiling water reactors;based on relevant data from the pressurized water Daya Bay nuclear power plant,a PWR core model is constructed with the Monte Carlo N-Particle Transport Code;this model suggests changes to existing fuel assemblies to enhance 60Co production.In addition,the plug rods are replaced with 59Co rods in the improved fuel assemblies in the simulation model to calculate critical parameters including the effective multiplication factor,neutron flux density,and distribution of energy deposition.By considering different numbers of 59Co rods,the simulation indicates that different layout schemes have different impact levels,but the impact is not large.As a whole,the components with four 59Co rods have a small impact,and the parameters of the reactor remain almost unchanged when four 59Co rods replace the secondary neutron source.Therefore,in theory,the use of a PWR to produce 60Co is feasible.展开更多
文摘One of the main issues in designing optimum tapered cascades for uranium enrichment for annual fuel production in a power reactor is whether to employ large(fat)or small(thin)cascades.What will be the permissible and optimal ranges of the number of machines that can be used in a cascade?For the first time,the permissible and optimal ranges of the number of gas centrifuges that can be utilized in a cascade were investigated using two types of centrifuges,and the performance of small and large tapered cascades was discussed.The particle swarm optimization algorithm(PSO)has been used to optimize tapered cascades.The results show:(1)For the first centrifuge,41 cascades(91≤n≤4897)and for the second centrifuge,49 cascades(18≤n≤3839)with small and large sizes can be used in enrichment facilities,and the best cascade for them has 530(with 23 stages)and 39(with 7 stages)centrifuges,respectively.(2)For both centrifuges,when 600≤n(number of centrifuges=n),the large cascade performance changes are relatively insignificant.(3)For both types of gas centrifuges,the annual los s of separation power in enrichment facilities is approximately 1.25%-4.82%of the total separation work required.
基金the National Natural Science Foundation of China(Grant No.U1967203)the National Key R&D Program of China(Grant No.2019YFB1901100)and China Postdoctoral Science Foundation(Grant No.2019M3737).
文摘Space nuclear reactor power(SNRP)using a gas-cooled reactor(GCR)and a closed Brayton cycle(CBC)is the ideal choice for future high-power space missions.To investigate the safety characteristics and develop the control strategies for gas-cooled SNRP,transient models for GCR,energy conversion unit,pipes,heat exchangers,pump and heat pipe radiator are established and a system analysis code is developed in this paper.Then,analyses of several operation conditions are performed using this code.In full-power steady-state operation,the core hot spot of 1293 K occurs near the upper part of the core.If 0.4$reactivity is introduced into the core,the maximum temperature that the fuel can reach is 2059 K,which is 914 K lower than the fuel melting point.The system finally has the ability to achieve a new steady-state with a higher reactor power.When the GCR is shut down in an emergency,the residual heat of the reactor can be removed through the conduction of the core and radiation heat transfer.The results indicate that the designed GCR is inherently safe owing to its negative reactivity feedback and passive decay heat removal.This paper may provide valuable references for safety design and analysis of the gas-cooled SNRP coupled with CBC.
文摘Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and economic operation for nuclear power plants( NPPs) especially for a large number of nuclear powers under construction. Based on the development and application of maintenance template in developed countries,and combining with reliability-centered maintenance( RCM) analysis results and maintenance experience data over the past ten years in domestic NPPs, the development process of maintenance template was presented for Chinese pressurized water reactor( PWR) NPP,and the application of maintenance template to maintenance program development and maintenance optimization combined with cases were demonstrated. A shortcut was provided for improving the efficiency of maintenance optimization in domestic PWR NPP,and help to realize a safe,reliable,and economic operation for domestic NPPs.
文摘It is very important to reduce the construction duration of the Reactor Containment Building (RCB) when considering the more than 50 months on average from concrete placement to completion. Through a case study, this study performs a pre-study for the reduction of construction duration in nuclear power plant project based on construction process of the RCB. The actual data of the case study have been collected and analyze the process and the external wall drawings of the RCB with construction practitioners. As a result of that, it is necessary to modularize the external wall form for equipment hatch and to extend the height of one layer of the external wall form to reduce the construction duration of RCB. The results of this study will be utilized to reduce construction duration of the nuclear power plant.
文摘This article looks for the necessary conditions to use Deuterium-Deuterium (D-D) fusion for a large power plant. At the moment, for nearly all the projects (JET, ITER…) only the Deuterium-Tritium (D-T) fuel is considered for a power plant. However, as shown in this article, even if a D-D reactor would be necessarily much bigger than a D-T reactor due to the much weaker fusion reactivity of the D-D fusion compared to the D-T fusion, a D-D reactor size would remain under an acceptable size. Indeed, a D-D power plant would be necessarily large and powerful, i.e. the net electric power would be equal to a minimum of 1.2 GWe and preferably above 10 GWe. A D-D reactor would be less complex than a D-T reactor as it is not necessary to obtain Tritium from the reactor itself. It is proposed the same type of reactor yet proposed by the author in a previous article, i.e. a Stellarator “racetrack” magnetic loop. The working of this reactor is continuous. It is reminded that the Deuterium is relatively abundant on the sea water, and so it constitutes an almost inexhaustible source of energy. Thanks to secondary fusions (D-T and D-He3) which both occur at an appreciable level above 100 keV, plasma can stabilize around such high equilibrium energy (i.e. between 100 and 150 keV). The mechanical gain (Q) of such reactor increases with the internal pipe radius, up to 4.5 m. A radius of 4.5 m permits a mechanical gain (Q) of about 17 which thanks to a modern thermo-dynamical conversion would lead to convert about 21% of the thermal power issued from the D-D reactor in a net electric power of 20 GWe. The goal of the article is to create a physical model of the D-D reactor so as to estimate this one without the need of a simulator and finally to estimate the dimensions, power and yield of such D-D reactor for different net electrical powers. The difficulties of the modeling of such reactor are listed in this article and would certainly be applicable to a future D-He3 reactor, if any.
文摘The paper describes the instrumentation for studying temperature and void reactivity effects that were developed at VR-I zero power reactor. Further are described its operational parameters, fields and ways of its utilization as well as issues connected to its implementation into the reactor core.
文摘This paper introduces the simulation, and controls using Simulink of MATLAB for PCTRAN (Personal Computer Transient Analysis) of the power control system (PWR) type pressurized water reactor of PWR WESTINGHOUSE AP1000. The power controller model produces mathematical model description in nonlinear relation form in Simulink of MATLAB which is an important and popular program used at most universities for education. The power controller is described by a block diagram in this paper and some details introduce to clearly understand the work function. The results of action control compared with the PCTRAN programme in modes of automatic and manual control.
文摘HPR1000 is an advanced nuclear power plant(NPP)with the significant feature of an active and passive safety design philosophy,developed by the China National Nuclear Corporation.On one hand,it is an evolutionary design based on proven technology of the existing pressurized water reactor NPP;on the other hand,it incorporates advanced design features including a 177-fuel-assembly core loaded with CF3 fuel assemblies,active and passive safety systems,comprehensive severe accident prevention and mitigation measures,enhanced protection against external events,and improved emergency response capability.Extensive verification experiments and tests have been performed for critical innovative improvements on passive systems,the reactor core,and the main equipment.The design of HPR1000fulfills the international utility requirements for advanced light water reactors and the latest nuclear safety requirements,and addresses the safety issues relevant to the Fukushima accident.Along with its outstanding safety and economy,HPR1000 provides an excellent and practicable solution for both domestic and international nuclear power markets.
基金supported by the National Natural Science Foundation of China(Nos.11635005 and 11705058)the Fundamental Research Funds for the Central Universities(No.2018ZD10)
文摘The radioactive isotope 60Co is used in many applications and is typically produced in heavy water reactors.As most of the commercial reactors in operation are pressurized light water reactors(PWRs),the world supply of high level radioactive cobalt would be greatly increased if 60Co could be produced in them.Currently,60Co production in PWRs has not been extensively studied;for the 59Co(n,c)60Co reaction,the positioning of 59Co rods in the reactor determines the rate of production.This article primarily uses the models of 60Co production in Canadian CANDU power reactors and American boiling water reactors;based on relevant data from the pressurized water Daya Bay nuclear power plant,a PWR core model is constructed with the Monte Carlo N-Particle Transport Code;this model suggests changes to existing fuel assemblies to enhance 60Co production.In addition,the plug rods are replaced with 59Co rods in the improved fuel assemblies in the simulation model to calculate critical parameters including the effective multiplication factor,neutron flux density,and distribution of energy deposition.By considering different numbers of 59Co rods,the simulation indicates that different layout schemes have different impact levels,but the impact is not large.As a whole,the components with four 59Co rods have a small impact,and the parameters of the reactor remain almost unchanged when four 59Co rods replace the secondary neutron source.Therefore,in theory,the use of a PWR to produce 60Co is feasible.
