The transfer of pressurized water reactor(PWR)technology from France to China is an important event in the history of Sino-French scientific and technological relations.China has gradually achieved self-reliance in th...The transfer of pressurized water reactor(PWR)technology from France to China is an important event in the history of Sino-French scientific and technological relations.China has gradually achieved self-reliance in the field of PWR technology through the introduction and subsequent absorption of France's 900 MW reactors.Compared with the process of introducing and absorbing similar technology from the United States by France,China's experience has been more complicated.This circumstance reflects the differences in the nuclear power technology systems between the two countries.France's industrial strength and early acquisition of nuclear power technology laid a solid foundation for mastering PWR technology.On the other hand,although China established a weak foundation through the implementation of the"728 Project,"and tried hard to negotiate with France,the substantive content of the technology transfer was very limited.By way of the policy transition from"unhooking of technology and trade"to"integration of technology and trade,"China ultimately accomplished the absorption and innovation of PWR technology through the Ling'ao NPP.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
Tracking load changes in a pressurized water reactor(PWR)with the help of an efficient core power control scheme in a nuclear power station is very important.The reason is that it is challenging to maintain a stable c...Tracking load changes in a pressurized water reactor(PWR)with the help of an efficient core power control scheme in a nuclear power station is very important.The reason is that it is challenging to maintain a stable core power according to the reference value within an acceptable tolerance for the safety of PWR.To overcome the uncertainties,a non-integer-based fractional order control method is demonstrated to control the core power of PWR.The available dynamic model of the reactor core is used in this analysis.Core power is controlled using a modified state feedback approach with a non-integer integral scheme through two different approximations,CRONE(Commande Robuste d’Ordre Non Entier,meaning Non-integer orderRobust Control)and FOMCON(non-integer order modeling and control).Simulation results are produced using MATLAB■program.Both non-integer results are compared with an integer order PI(Proportional Integral)algorithm to justify the effectiveness of the proposed scheme.Sate-spacemodel Core power control Non-integer control Pressurized water reactor PI controller CRONE FOMCON.展开更多
In resonance with the Fukushima Daiichi Nuclear Power Plant accident lesson, a novel fuel design to enhance safety regarding severe accident scenarios has become increasingly appreciated in the nuclear power industry....In resonance with the Fukushima Daiichi Nuclear Power Plant accident lesson, a novel fuel design to enhance safety regarding severe accident scenarios has become increasingly appreciated in the nuclear power industry. This research focuses on analysis of the neutronic properties of a silicon carbide(SiC) cladding fuel assembly, which provides a greater safety margin as a type of accident-tolerant fuel for pressurized water reactors. The general physical performance of SiC cladding is explored to ascertain its neutronic performance. The neutron spectrum, accumulation of ^(239)Pu, physical characteristics,temperature reactivity coefficient, and power distribution are analyzed. Furthermore, the influences of a burnable poison rod and enrichment are explored. SiC cladding assemblies show a softer neutron spectrum and flatter power distribution than conventional Zr alloy cladding fuel assemblies. Lower enrichment fuel is required when SiC cladding is adopted. However, the positive reactivity coefficient associated with the SiC material remains to be offset. The results reveal that SiC cladding assemblies show broad agreement with the neutronic performance of conventional Zr alloy cladding fuel. In the meantime, its unique physical characteristics can lead to improved safety and economy.展开更多
This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single...This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the pre- liminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the pri- mary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation.展开更多
In this paper, response surface methodology(RSM) based on central composite design(CCD) is applied to obtain an optimization design for the fuel rod's diameter and distance cooled by turbulent Al_2O_3–water nanof...In this paper, response surface methodology(RSM) based on central composite design(CCD) is applied to obtain an optimization design for the fuel rod's diameter and distance cooled by turbulent Al_2O_3–water nanofluid for a typical pressurized water reactor(PWR). Fuel rods and nanofluid flow between them are simulated 3D using computational fluid dynamics(CFD) by ANSYS-FLUNET package software. The RNG k–ε model is used to simulate turbulent nanofluid flow between the rods. The effect of different nanoparticles concentration is also investigated on the Nusselt number from heat transfer efficiency view point. Results reveal that when distance parameter(a) is in the minimum level and diameter parameter(r) is in the maximum possible level, cooling the rods will be better due to higher Nusselt number in this situation. Also, using the different nanoparticles on the cooling process confirms that Al_2O_3 averagely 17% and TiO_2 10% improve the Nusselt numbers.展开更多
The design of an integral pressurized water reactor (IPWR) focuses on enhancing the safety and reliability of the reactor by incorporating a number of inherent safety features and engineered safety features. However, ...The design of an integral pressurized water reactor (IPWR) focuses on enhancing the safety and reliability of the reactor by incorporating a number of inherent safety features and engineered safety features. However, the characteristics of passive safety systems for the marine reactors are quiet different from those for the land nuclear power plant because of the more formidable and dangerous operation environments of them. This paper presents results of marine black out accident analyses. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the RELAP5/MOD3.4 code and a sensitivity analysis has been performed to evaluate a passive cooldown capability. The results of the accident analyses show that the reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation.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
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
Stress corrosion cracking(SCC) of an SA508-309 L/308 L–316 L dissimilar metal weld joint in primary pressurized water reactor environment was investigated by the interrupted slow strain rate tension tests following a...Stress corrosion cracking(SCC) of an SA508-309 L/308 L–316 L dissimilar metal weld joint in primary pressurized water reactor environment was investigated by the interrupted slow strain rate tension tests following a microstructure characterization. The 308 L weld metal shows a higher content of δ ferrite than the 309 L weld metal. In addition, no obvious Cr-depletion but carbides precipitation at δ phase boundaries was observed in both 308 L and 309 L weld metals. The slow strain rate tension tests showed that the SCC susceptibility of the base and weld metals of the dissimilar metal weld joint follows the order of SA508 < 308 L weld metal < the heat affected zone of 316 L base metal < 309 L weld metal.The higher SCC susceptibility of 309 L weld metal than that of 308 L weld metal is likely due to the lower content of δ ferrite. In addition, a preferential SCC initiation in the 309 L weld metal adjacent to 308 L weld metal is attributed to few carbides in this region.展开更多
The WRNM(wide range neutron monitoring)is a newly developed neutron monitoring channel which was initially conceived as a means to meet Regulatory Guide 1.97 requirements for post-accident neutron monitoring.The scope...The WRNM(wide range neutron monitoring)is a newly developed neutron monitoring channel which was initially conceived as a means to meet Regulatory Guide 1.97 requirements for post-accident neutron monitoring.The scope was expanded to include the startup monitoring function with the aim of replacing both the source and IRMs(intermediate range monitors)in BWRs(boiling water reactors).The WRNMs,consisting of a newly designed fixed incore regenerative sensor and new electronics,which include both counting and MSV(mean square voltage)channels,have been tested in several reactors and its capabilities have been confirmed.The channel will cover the neutron flux range from 103 nv to 1.5×103 nv;it has greater than 1 decade overlap between the counting and MSV channels.Because of the regenerative fissile coating the sensor,even though fixed incore,has a life of approximately 6.0 full power years in a 51 kW/L BWR and similar situation has been proposed for newly designed small modular reactor such as BWRX-300 of General Electric Hitachi reactor.展开更多
To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up, a tight pitch lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors. It is shown t...To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up, a tight pitch lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors. It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs. Various techniques were proposed to solve these problems. In this work, a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated. By utilizing numerical simulation technique, it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio (0.98), long burn-up (60 GWD/t) and negative void reactivity coefficients.展开更多
Small-break superposed station blackout(SBO)accidents are the basic design accidents of nuclear power plants.Under the condition of a small break in the cold leg,SBO further increases the severity of the accident,and ...Small-break superposed station blackout(SBO)accidents are the basic design accidents of nuclear power plants.Under the condition of a small break in the cold leg,SBO further increases the severity of the accident,and the steam bypass discharg-ing system(GCT)in the second circuit can play an important role in guaranteeing core safety.To explore the influence of the GCT on the thermal-hydraulic characteristics of the primary circuit,RELAP5 software was used to establish a numerical model based on a typical pressurized water reactor nuclear power plant.Five different small breaks in the cold-leg super-posed SBO were selected,and the impact of the GCT operation on the transient response characteristics of the primary and secondary circuit systems was analyzed.The results show that the GCT plays an indispensable role in core heat removal during an accident;otherwise,core safety cannot be guaranteed.The GCT was used in conjunction with the primary safety injection system during the placement process.When the break diameter was greater than a certain critical value,the core cooling rate could not be guaranteed to be less than 100 K/h;however,the core remained in a safe state.展开更多
An in-house code,CONTHAC-3D,was developed to calculate and analyze thermal-hydraulic phenomena in containments during severe accidents.CONTHAC-3D is a three-dimensional computational fluid dynamics code that can be ap...An in-house code,CONTHAC-3D,was developed to calculate and analyze thermal-hydraulic phenomena in containments during severe accidents.CONTHAC-3D is a three-dimensional computational fluid dynamics code that can be applied to predict gas flow,diffusion,and steam condensation in a containment during a severe hypothetical accident,as well as to obtain an estimate of the local hydrogen concentration in various zones of the containment.CONTHAC-3D was developed using multiple models to simulate the features of the proprietary systems and equipment of HPR1000 and ACP100,such as the passive cooling system,passive autocatalytic recombiners and the passive air cooling system.To validate CONTHAC-3D,a GX6 test was performed at the Battelle Model Containment facility.The hydrogen concentration and temperature monitored by the GX6 test are accurately predicted by CONTHAC-3D.Subsequently,the hydrogen distribution in the HPR1000 containment during a severe accident was studied.The results show that the hydrogen removal rates calculated using CONTHAC-3D for different types of PARs agree well with the theoretical values,with an error of less than 1%.As the accident progresses,the hydrogen concentration in the lower compartment becomes higher than that in the large space,which implies that the lower compartment has a higher hydrogen risk than the dome and large space at a later stage of the accident.The amount of hydrogen removed by the PARs placed on the floor of the compartment is small;therefore,raising the installation height of these recombiners appropriately is recommended.However,we do not recommend installing all autocatalytic recombiners at high positions.The study findings in regard to the hydrogen distribution in the HPR1000 containment indicate that CONTHAC-3D can be applied to the study of hydrogen risk containment.展开更多
基金a phase study of a key project of the Fourteenth Five-Year Plan of the Institute for the History of Natural Sciences,Chinese Academy of Sciences:“A Comparative Study of the Sino-Foreign History of Scientific and Technological Innovation:The Road to Scientific and Technological Self-Reliance and Self-Improvement”,E2291J01。
文摘The transfer of pressurized water reactor(PWR)technology from France to China is an important event in the history of Sino-French scientific and technological relations.China has gradually achieved self-reliance in the field of PWR technology through the introduction and subsequent absorption of France's 900 MW reactors.Compared with the process of introducing and absorbing similar technology from the United States by France,China's experience has been more complicated.This circumstance reflects the differences in the nuclear power technology systems between the two countries.France's industrial strength and early acquisition of nuclear power technology laid a solid foundation for mastering PWR technology.On the other hand,although China established a weak foundation through the implementation of the"728 Project,"and tried hard to negotiate with France,the substantive content of the technology transfer was very limited.By way of the policy transition from"unhooking of technology and trade"to"integration of technology and trade,"China ultimately accomplished the absorption and innovation of PWR technology through the Ling'ao NPP.
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.
基金This project was funded by the Deanship of Scientific Research(DSR),King Abdulaziz University,Jeddah,Saudi Arabia under grant no.(KEP-Msc-36-135-38).
文摘Tracking load changes in a pressurized water reactor(PWR)with the help of an efficient core power control scheme in a nuclear power station is very important.The reason is that it is challenging to maintain a stable core power according to the reference value within an acceptable tolerance for the safety of PWR.To overcome the uncertainties,a non-integer-based fractional order control method is demonstrated to control the core power of PWR.The available dynamic model of the reactor core is used in this analysis.Core power is controlled using a modified state feedback approach with a non-integer integral scheme through two different approximations,CRONE(Commande Robuste d’Ordre Non Entier,meaning Non-integer orderRobust Control)and FOMCON(non-integer order modeling and control).Simulation results are produced using MATLAB■program.Both non-integer results are compared with an integer order PI(Proportional Integral)algorithm to justify the effectiveness of the proposed scheme.Sate-spacemodel Core power control Non-integer control Pressurized water reactor PI controller CRONE FOMCON.
基金supported by the National Natural Science Foundation of China(No.11675057)the Fundamental Research Funds for the Central Universities(No.2017ZD100)
文摘In resonance with the Fukushima Daiichi Nuclear Power Plant accident lesson, a novel fuel design to enhance safety regarding severe accident scenarios has become increasingly appreciated in the nuclear power industry. This research focuses on analysis of the neutronic properties of a silicon carbide(SiC) cladding fuel assembly, which provides a greater safety margin as a type of accident-tolerant fuel for pressurized water reactors. The general physical performance of SiC cladding is explored to ascertain its neutronic performance. The neutron spectrum, accumulation of ^(239)Pu, physical characteristics,temperature reactivity coefficient, and power distribution are analyzed. Furthermore, the influences of a burnable poison rod and enrichment are explored. SiC cladding assemblies show a softer neutron spectrum and flatter power distribution than conventional Zr alloy cladding fuel assemblies. Lower enrichment fuel is required when SiC cladding is adopted. However, the positive reactivity coefficient associated with the SiC material remains to be offset. The results reveal that SiC cladding assemblies show broad agreement with the neutronic performance of conventional Zr alloy cladding fuel. In the meantime, its unique physical characteristics can lead to improved safety and economy.
文摘This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the pre- liminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the pri- mary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation.
基金financial support of the National Natural Science Foundation of China (No. 51422604, 21276206)the National 863 Program of China (No. 2013AA050402)supported by the China Fundamental Research Funds for the Central Universities
文摘In this paper, response surface methodology(RSM) based on central composite design(CCD) is applied to obtain an optimization design for the fuel rod's diameter and distance cooled by turbulent Al_2O_3–water nanofluid for a typical pressurized water reactor(PWR). Fuel rods and nanofluid flow between them are simulated 3D using computational fluid dynamics(CFD) by ANSYS-FLUNET package software. The RNG k–ε model is used to simulate turbulent nanofluid flow between the rods. The effect of different nanoparticles concentration is also investigated on the Nusselt number from heat transfer efficiency view point. Results reveal that when distance parameter(a) is in the minimum level and diameter parameter(r) is in the maximum possible level, cooling the rods will be better due to higher Nusselt number in this situation. Also, using the different nanoparticles on the cooling process confirms that Al_2O_3 averagely 17% and TiO_2 10% improve the Nusselt numbers.
文摘The design of an integral pressurized water reactor (IPWR) focuses on enhancing the safety and reliability of the reactor by incorporating a number of inherent safety features and engineered safety features. However, the characteristics of passive safety systems for the marine reactors are quiet different from those for the land nuclear power plant because of the more formidable and dangerous operation environments of them. This paper presents results of marine black out accident analyses. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the RELAP5/MOD3.4 code and a sensitivity analysis has been performed to evaluate a passive cooldown capability. The results of the accident analyses show that the reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation.
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.
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.
基金supported by the National Natural Science Foundation of China(No.51571204)the support of the Youth Scientific and Innovation Research Team for Advanced Surface Functional Materials(Southwest Petroleum University,No.2018CXTD06)the Young Scholars Development Found of Southwest Petroleum University(No.201899010040).
文摘Stress corrosion cracking(SCC) of an SA508-309 L/308 L–316 L dissimilar metal weld joint in primary pressurized water reactor environment was investigated by the interrupted slow strain rate tension tests following a microstructure characterization. The 308 L weld metal shows a higher content of δ ferrite than the 309 L weld metal. In addition, no obvious Cr-depletion but carbides precipitation at δ phase boundaries was observed in both 308 L and 309 L weld metals. The slow strain rate tension tests showed that the SCC susceptibility of the base and weld metals of the dissimilar metal weld joint follows the order of SA508 < 308 L weld metal < the heat affected zone of 316 L base metal < 309 L weld metal.The higher SCC susceptibility of 309 L weld metal than that of 308 L weld metal is likely due to the lower content of δ ferrite. In addition, a preferential SCC initiation in the 309 L weld metal adjacent to 308 L weld metal is attributed to few carbides in this region.
文摘The WRNM(wide range neutron monitoring)is a newly developed neutron monitoring channel which was initially conceived as a means to meet Regulatory Guide 1.97 requirements for post-accident neutron monitoring.The scope was expanded to include the startup monitoring function with the aim of replacing both the source and IRMs(intermediate range monitors)in BWRs(boiling water reactors).The WRNMs,consisting of a newly designed fixed incore regenerative sensor and new electronics,which include both counting and MSV(mean square voltage)channels,have been tested in several reactors and its capabilities have been confirmed.The channel will cover the neutron flux range from 103 nv to 1.5×103 nv;it has greater than 1 decade overlap between the counting and MSV channels.Because of the regenerative fissile coating the sensor,even though fixed incore,has a life of approximately 6.0 full power years in a 51 kW/L BWR and similar situation has been proposed for newly designed small modular reactor such as BWRX-300 of General Electric Hitachi reactor.
文摘To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up, a tight pitch lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors. It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs. Various techniques were proposed to solve these problems. In this work, a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated. By utilizing numerical simulation technique, it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio (0.98), long burn-up (60 GWD/t) and negative void reactivity coefficients.
文摘Small-break superposed station blackout(SBO)accidents are the basic design accidents of nuclear power plants.Under the condition of a small break in the cold leg,SBO further increases the severity of the accident,and the steam bypass discharg-ing system(GCT)in the second circuit can play an important role in guaranteeing core safety.To explore the influence of the GCT on the thermal-hydraulic characteristics of the primary circuit,RELAP5 software was used to establish a numerical model based on a typical pressurized water reactor nuclear power plant.Five different small breaks in the cold-leg super-posed SBO were selected,and the impact of the GCT operation on the transient response characteristics of the primary and secondary circuit systems was analyzed.The results show that the GCT plays an indispensable role in core heat removal during an accident;otherwise,core safety cannot be guaranteed.The GCT was used in conjunction with the primary safety injection system during the placement process.When the break diameter was greater than a certain critical value,the core cooling rate could not be guaranteed to be less than 100 K/h;however,the core remained in a safe state.
文摘An in-house code,CONTHAC-3D,was developed to calculate and analyze thermal-hydraulic phenomena in containments during severe accidents.CONTHAC-3D is a three-dimensional computational fluid dynamics code that can be applied to predict gas flow,diffusion,and steam condensation in a containment during a severe hypothetical accident,as well as to obtain an estimate of the local hydrogen concentration in various zones of the containment.CONTHAC-3D was developed using multiple models to simulate the features of the proprietary systems and equipment of HPR1000 and ACP100,such as the passive cooling system,passive autocatalytic recombiners and the passive air cooling system.To validate CONTHAC-3D,a GX6 test was performed at the Battelle Model Containment facility.The hydrogen concentration and temperature monitored by the GX6 test are accurately predicted by CONTHAC-3D.Subsequently,the hydrogen distribution in the HPR1000 containment during a severe accident was studied.The results show that the hydrogen removal rates calculated using CONTHAC-3D for different types of PARs agree well with the theoretical values,with an error of less than 1%.As the accident progresses,the hydrogen concentration in the lower compartment becomes higher than that in the large space,which implies that the lower compartment has a higher hydrogen risk than the dome and large space at a later stage of the accident.The amount of hydrogen removed by the PARs placed on the floor of the compartment is small;therefore,raising the installation height of these recombiners appropriately is recommended.However,we do not recommend installing all autocatalytic recombiners at high positions.The study findings in regard to the hydrogen distribution in the HPR1000 containment indicate that CONTHAC-3D can be applied to the study of hydrogen risk containment.