This study investigated the feasibility of distance protection in extra-high voltage(EHV) networks. In long-distance transmission lines, the distributed parameter characteristic of the EHV network is obvious. When a f...This study investigated the feasibility of distance protection in extra-high voltage(EHV) networks. In long-distance transmission lines, the distributed parameter characteristic of the EHV network is obvious. When a fault occurs far away from the measurement site, the measured impedance might not be directly proportional to the fault distance, and the protection domain of distance protection will be decreased. The detailed theory inferred and proven in this paper reveals that this phenomenon is widespread in EHV transmission lines. The results indicate that the protection domain error is greatly reduced by the application of the shunt reactor. Overall, simulation results show that the proposed method is effective for impedance relay, considering different characteristics, different lengths of lines, and compensation degrees.展开更多
The digital reactor protection system(RPS)is one of the most important digital instrumentation and control(I&C)systems utilized in nuclear power plants(NPPs).It ensures a safe reactor trip when the safety-related ...The digital reactor protection system(RPS)is one of the most important digital instrumentation and control(I&C)systems utilized in nuclear power plants(NPPs).It ensures a safe reactor trip when the safety-related parameters violate the operational limits and conditions of the reactor.Achieving high reliability and availability of digital RPS is essential to maintaining a high degree of reactor safety and cost savings.The main objective of this study is to develop a general methodology for improving the reliability of the RPS in NPP,based on a Bayesian Belief Network(BBN)model.The structure of BBN models is based on the incorporation of failure probability and downtime of the RPS I&C components.Various architectures with dual-state nodes for the I&C components were developed for reliability-sensitive analysis and availability optimization of the RPS and to demonstrate the effect of I&C components on the failure of the entire system.A reliability framework clarified as a reliability block diagram transformed into a BBN representation was constructed for each architecture to identify which one will fit the required reliability.The results showed that the highest availability obtained using the proposed method was 0.9999998.There are 120 experiments using two common component importance measures that are applied to define the impact of I&C modules,which revealed that some modules are more risky than others and have a larger effect on the failure of the digital RPS.展开更多
Safety system testing is one of the most rigorous and time-consuming requirements in the verification and validation process for reactor protection systems(RPSs).This paper presents the development of a test system fo...Safety system testing is one of the most rigorous and time-consuming requirements in the verification and validation process for reactor protection systems(RPSs).This paper presents the development of a test system for the fully digital and field-programmable gate array-based RPS of the solid fuel(SF) thorium-breeding molten salt pebble bed fluoride salt-cooled reactor(TMSR),denoted as the TMSR-SF1 project,developed by the Chinese Academy of Sciences.The test system is applied to the RPS to ensure that it fully meets its designed functions and system specifications.We first introduce the testing principles and methods.Then,the hardware component designs and the software program development of the test system are discussed.Finally,the test process and test results are discussed and summarized.展开更多
Fast reactors used lead-bismuth eutectic (LBE) and lead as coolants possess very high level of inherent self-protection and passive safety against severe accident. So, population radiophobia can be overcome. That type...Fast reactors used lead-bismuth eutectic (LBE) and lead as coolants possess very high level of inherent self-protection and passive safety against severe accident. So, population radiophobia can be overcome. That type of reactors can be simultaneously more safely and more cheaply. As all other coolants, LBE and lead coolant (LC) possess the certain virtues and shortcomings. The presented report includes the comparative analysis of characteristic properties of those coolants, their impact on reactor safety, reliability and operating characteristics. The conclusion is made about promising usage of FRs with these coolants in future NP after the experience in operating of the prototypes of such reactors has been obtained.展开更多
The fluoride volatility method (FVM) is a technique tailored to separate uranium from fuel salt of molten salt reactors. A key challenge in R&D of the FVM is corrosion due to the presence of molten salt and corros...The fluoride volatility method (FVM) is a technique tailored to separate uranium from fuel salt of molten salt reactors. A key challenge in R&D of the FVM is corrosion due to the presence of molten salt and corrosive gases at high temperature. In this work, a frozen-wall technique was proposed to produce a physical barrier between construction materials and corrosive reactants. The protective performance of the frozen wall against molten salt was assessed using FLiNaK molten salt with introduced fluorine gas, which was regarded as a simulation of the FVM process. SS304, SS316L, Inconel 600 and graphite were chosen as the test samples. The extent of corrosion was characterized by an analysis of weight loss and scanning electron microscope studies. All four test samples suffered severe corrosion in the molten salt phase with the corrosion resistance as: Inconel 600>SS316L>graphite>SS304. The presence of the frozen wall could protect materials against corrosion by molten salt and corrosive gases, and compared with materials exposed to molten salt, the corrosion rates of materials protected by the frozen wall were decreased by at least one order of magnitude.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
文摘This study investigated the feasibility of distance protection in extra-high voltage(EHV) networks. In long-distance transmission lines, the distributed parameter characteristic of the EHV network is obvious. When a fault occurs far away from the measurement site, the measured impedance might not be directly proportional to the fault distance, and the protection domain of distance protection will be decreased. The detailed theory inferred and proven in this paper reveals that this phenomenon is widespread in EHV transmission lines. The results indicate that the protection domain error is greatly reduced by the application of the shunt reactor. Overall, simulation results show that the proposed method is effective for impedance relay, considering different characteristics, different lengths of lines, and compensation degrees.
文摘The digital reactor protection system(RPS)is one of the most important digital instrumentation and control(I&C)systems utilized in nuclear power plants(NPPs).It ensures a safe reactor trip when the safety-related parameters violate the operational limits and conditions of the reactor.Achieving high reliability and availability of digital RPS is essential to maintaining a high degree of reactor safety and cost savings.The main objective of this study is to develop a general methodology for improving the reliability of the RPS in NPP,based on a Bayesian Belief Network(BBN)model.The structure of BBN models is based on the incorporation of failure probability and downtime of the RPS I&C components.Various architectures with dual-state nodes for the I&C components were developed for reliability-sensitive analysis and availability optimization of the RPS and to demonstrate the effect of I&C components on the failure of the entire system.A reliability framework clarified as a reliability block diagram transformed into a BBN representation was constructed for each architecture to identify which one will fit the required reliability.The results showed that the highest availability obtained using the proposed method was 0.9999998.There are 120 experiments using two common component importance measures that are applied to define the impact of I&C modules,which revealed that some modules are more risky than others and have a larger effect on the failure of the digital RPS.
基金supported by the Leading Science and Technology Project of Chinese Academy of Sciences(No.XD02001003)
文摘Safety system testing is one of the most rigorous and time-consuming requirements in the verification and validation process for reactor protection systems(RPSs).This paper presents the development of a test system for the fully digital and field-programmable gate array-based RPS of the solid fuel(SF) thorium-breeding molten salt pebble bed fluoride salt-cooled reactor(TMSR),denoted as the TMSR-SF1 project,developed by the Chinese Academy of Sciences.The test system is applied to the RPS to ensure that it fully meets its designed functions and system specifications.We first introduce the testing principles and methods.Then,the hardware component designs and the software program development of the test system are discussed.Finally,the test process and test results are discussed and summarized.
文摘Fast reactors used lead-bismuth eutectic (LBE) and lead as coolants possess very high level of inherent self-protection and passive safety against severe accident. So, population radiophobia can be overcome. That type of reactors can be simultaneously more safely and more cheaply. As all other coolants, LBE and lead coolant (LC) possess the certain virtues and shortcomings. The presented report includes the comparative analysis of characteristic properties of those coolants, their impact on reactor safety, reliability and operating characteristics. The conclusion is made about promising usage of FRs with these coolants in future NP after the experience in operating of the prototypes of such reactors has been obtained.
基金supported by the Strategic Priority Research Program of the Chinese Academy of Science(No.XDA02030000)
文摘The fluoride volatility method (FVM) is a technique tailored to separate uranium from fuel salt of molten salt reactors. A key challenge in R&D of the FVM is corrosion due to the presence of molten salt and corrosive gases at high temperature. In this work, a frozen-wall technique was proposed to produce a physical barrier between construction materials and corrosive reactants. The protective performance of the frozen wall against molten salt was assessed using FLiNaK molten salt with introduced fluorine gas, which was regarded as a simulation of the FVM process. SS304, SS316L, Inconel 600 and graphite were chosen as the test samples. The extent of corrosion was characterized by an analysis of weight loss and scanning electron microscope studies. All four test samples suffered severe corrosion in the molten salt phase with the corrosion resistance as: Inconel 600>SS316L>graphite>SS304. The presence of the frozen wall could protect materials against corrosion by molten salt and corrosive gases, and compared with materials exposed to molten salt, the corrosion rates of materials protected by the frozen wall were decreased by at least one order of magnitude.
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.