Aiming to meet the demand of the country＇ s nuclear demonstration project on the CAP1400 nuclear power plant, Baosteel uses the roll-bonding technology and develops the SA-533 Type B CL. 1 ＋ SA-240 Type 304L high-st...Aiming to meet the demand of the country＇ s nuclear demonstration project on the CAP1400 nuclear power plant, Baosteel uses the roll-bonding technology and develops the SA-533 Type B CL. 1 ＋ SA-240 Type 304L high-strength and high-toughness clad steel plate with a shear strength of over 310 MPa for the nuclear power plant＇ s safety injection tank. The properties of the quenched and tempered and the simulated post-weld heat treatment states are systematically studied herein through a comprehensive inspection and evaluation of the composition,microstructure,and properties of the clad steel plate. The results show that the bonding interface has high shear strength and that the base metal has high strength and good toughness at low temperatures. Hence, the performance fully meets the technical requirements of the CAP1400 nuclear power plant＇ s safety injection tank in the country＇ s nuclear demonstration project. The roll-bonded clad steel plate can be used to manufacture the safety injection tank of the CAP1400 nuclear power plant.展开更多
Abstract： The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete...Abstract： The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel （PCCV） model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.展开更多
A nuclear power plant real-time engineering simulator was developed based on general-purpose ther- mal-hydraulic system simulation code RELAP5. It mainly consists of three parts:improved thermal-hydraulic system simul...A nuclear power plant real-time engineering simulator was developed based on general-purpose ther- mal-hydraulic system simulation code RELAP5. It mainly consists of three parts:improved thermal-hydraulic system simulation code RELAP5,control and protection system and human-machine interface. A normal transient of CHASHMA nuclear power plant turbine step load change from 100% to 90% of full power,was simulated by the en- gineering simulator as an application example. This paper presents structure and main features of the engineering simulator,and application results are shown and discussed.展开更多
The original internal flooding probabilistic safety analysis （PSA） study of Krsko Nuclear Power Plant （two-loop Pressurized Water Reactor （PWR） plant of Westinghouse design） was performed in mid nineties and lim...The original internal flooding probabilistic safety analysis （PSA） study of Krsko Nuclear Power Plant （two-loop Pressurized Water Reactor （PWR） plant of Westinghouse design） was performed in mid nineties and limited to reactor core damage risk （Level 1 PSA）. In 2003, it was, together with other safety and hazard analyses, subject to the Periodic Safety Review （PSR）. In the PSR, it was stated that methodological PSA approaches and guidelines have evoluted during the past decade and several observations were provided, concerning the area screening process, residual risk and treatment of plant damage states and risk from radioactivity releases （i.e., Level 2 PSA）. In order to address the PSR observations, upgrade ofKrsko NPP internal flooding PSA was undertaken. The area screening process was revisited in order to cover the areas without automatic reactor trip equipment. The model was extended to Level 2. Residual risk was estimated at both Level 1 and Level 2, in terms of core damage frequency （CDF） and large early release frequency （LERF）, respectively.展开更多
This paper presents main results of a comprehensive study performed by the European Clearinghouse on Operating Experience Feedback of NPP with the support of IRSN (Institut de S?reté Nucléaire et de Radiopro...This paper presents main results of a comprehensive study performed by the European Clearinghouse on Operating Experience Feedback of NPP with the support of IRSN (Institut de S?reté Nucléaire et de Radioprotection) and GRS (Gesellschaft für Anlagen und Reaktorsi-cherheit mbH). Physical ageing mechanisms of structure, systems and components that eventually lead to ageing related systems and components failures at nuclear power plants were the main focus of this study. The analysis of ageing related events involved operating experience reported by NPP operators in France, Germany, USA and to the IAEA/NEA International Reporting system, on operating experience for the past 20 years (i.e. 1990-2009). A list of ageing related events was populated. Each ageing related event contained in the list was analyzed and results of analysis were summarized for each commodity group for which the ageing degradation appeared to be a dominant contributor or direct cause. The most common degradation mechanisms/ageing effects for each specific component/commodity group, their risk significance and consequences to the plant performance are described. This paper provides insights into ageing related operating experience as well as recommendations to deal with the physical ageing of nuclear power plant SSC important to safety.展开更多
The conflict between safety and economics requirements is peculiar to the present nuclear power (NP). The main point of the conflict is that for traditional type reactors the increase of requirements to safety of nucl...The conflict between safety and economics requirements is peculiar to the present nuclear power (NP). The main point of the conflict is that for traditional type reactors the increase of requirements to safety of nuclear power plants (NPP) worsens their economical characteristics. This is caused by large potential energy accumulated in reactor coolant. In the presented paper the opportunity and expediency of changeover to reactors with heavy liquid-metal coolants (HLMC) in future NP is grounded. First of all, this refers to lead-bismuth coolant (LBC) mastered in the process of operating nuclear submarines (NS) reactors. The reactor facilities (RFs) of that type cannot cause destruction of defense barriers and make possible deterministic elimination of severe accidents with catastrophic radioactivity release. So it will make possible to eliminate the highlighted conflict and reasons for existence of population’s radiophobia. Lead-bismuth fast reactor SVBR-100 with electric power of 100 MWe is the reactor facility of that type. The effect of accumulated in coolant potential energy on safety and economics is considered. Main specific features of SVBR-100 technology providing a high level of inherent self-protection and passive safety are presented.展开更多
Polyethylene （PE） pipe, particularly high- density polyethylene （HDPE） pipe, has been successfully utilized to transport cooling water for both non-safety- and safety-related applications in nuclear power plant （...Polyethylene （PE） pipe, particularly high- density polyethylene （HDPE） pipe, has been successfully utilized to transport cooling water for both non-safety- and safety-related applications in nuclear power plant （NPP）. Though ASME Code Case N755, which is the first code case related to NPP HDPE pipe, requires a thorough nondestructive examination （NDE） of HDPE joints. However, no executable regulations presently exist because of the lack of a feasible NDE technique for HDPE pipe in NPP. This work presents a review of current developments in NDE technology for both HDPE pipe in NPP with a diameter of less than 400 mm and that of a larger size. For the former category, phased array ultrasonic technique is proven effective for inspecting typical defects in HDPE pipe, and is thus used in Chinese national standards GB/T 29460 and GB/T 29461. A defect- recognition technique is developed based on pattern recognition, and a safety assessment principle is summa- rized from the database of destructive testing. On the other hand, recent research and practical studies reveal that in current ultrasonic-inspection technology, the absence of effective ultrasonic inspection for large size was lack of consideration of the viscoelasticity effect of PE on acoustic wave propagation in current ultrasonic inspection technology. Furthermore, main technical problems were analyzed in the paper to achieve an effective ultrasonic test method in accordance to the safety and efficiency requirements of related regulations and standards. Finally, the development trend and challenges of NDE test technology for HDPE in NPP are discussed.展开更多
Many developing countries need ecologically clean power sources (PS). The nuclear power plants are such sources. However, a great number of the developing countries do not possess developed large capacity power system...Many developing countries need ecologically clean power sources (PS). The nuclear power plants are such sources. However, a great number of the developing countries do not possess developed large capacity power systems. Moreover, currently in the developing countries, there are no highly skilled personnel to provide construction and reliable and safe operation of the nuclear plants, which are complex and potentially hazardous systems. In some countries, the level of terroristic threat is extremely high. For that reason, there are specific requirements to the nuclear PSs intended for use in the developing countries. In the presented report, the specific requirements which must be met by the NPT proposed for use in developing countries are formulated, basic statements of the SVBR-100 concept are presented, design and principal scheme of the reactor fa-ility are described, major characteristics of SVBR-100 are summarized.展开更多
This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behavior at shutdown reactor during maintenance. For the purpose of the analysis is selected a plant operating s...This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behavior at shutdown reactor during maintenance. For the purpose of the analysis is selected a plant operating state with unsealed primary circuit by removing the MCP head. The reference nuclear power plant is Unit 6 at Kozloduy NPP (KNPP) site. RELAP5/ MOD3.2 computer code has been used to simulate the transient for WWER-1000/V320 NPP model. A model of WWER-1000 based on Unit 6 of KNPP has been developed for the RELAP5/MOD3.2 code at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS), Sofia. The plant modifications performed in frame of modernization program have been taken into account for the investigated conditions for the unsealed primary circuit. The most specific in this analysis compared to the analyses of NPP accidents at full power is the unavailability of some important safety systems. For the purpose of the present investigation two scenarios have been studied, involving a different number of safety systems with and without operator actions. The selected initiating event and scenarios are used in support of analytical validation of Emergency Operating Procedures (EOP) at low power and they are based on the suggestions of leading KNPP experts and are important in support of analytical validation of EOP at low power.展开更多
文摘Aiming to meet the demand of the country＇ s nuclear demonstration project on the CAP1400 nuclear power plant, Baosteel uses the roll-bonding technology and develops the SA-533 Type B CL. 1 ＋ SA-240 Type 304L high-strength and high-toughness clad steel plate with a shear strength of over 310 MPa for the nuclear power plant＇ s safety injection tank. The properties of the quenched and tempered and the simulated post-weld heat treatment states are systematically studied herein through a comprehensive inspection and evaluation of the composition,microstructure,and properties of the clad steel plate. The results show that the bonding interface has high shear strength and that the base metal has high strength and good toughness at low temperatures. Hence, the performance fully meets the technical requirements of the CAP1400 nuclear power plant＇ s safety injection tank in the country＇ s nuclear demonstration project. The roll-bonded clad steel plate can be used to manufacture the safety injection tank of the CAP1400 nuclear power plant.
基金National Natural Science Foundation of China under Grant Nos.51138001 and 51479027
文摘Abstract： The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel （PCCV） model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.
文摘A nuclear power plant real-time engineering simulator was developed based on general-purpose ther- mal-hydraulic system simulation code RELAP5. It mainly consists of three parts:improved thermal-hydraulic system simulation code RELAP5,control and protection system and human-machine interface. A normal transient of CHASHMA nuclear power plant turbine step load change from 100% to 90% of full power,was simulated by the en- gineering simulator as an application example. This paper presents structure and main features of the engineering simulator,and application results are shown and discussed.
文摘The original internal flooding probabilistic safety analysis （PSA） study of Krsko Nuclear Power Plant （two-loop Pressurized Water Reactor （PWR） plant of Westinghouse design） was performed in mid nineties and limited to reactor core damage risk （Level 1 PSA）. In 2003, it was, together with other safety and hazard analyses, subject to the Periodic Safety Review （PSR）. In the PSR, it was stated that methodological PSA approaches and guidelines have evoluted during the past decade and several observations were provided, concerning the area screening process, residual risk and treatment of plant damage states and risk from radioactivity releases （i.e., Level 2 PSA）. In order to address the PSR observations, upgrade ofKrsko NPP internal flooding PSA was undertaken. The area screening process was revisited in order to cover the areas without automatic reactor trip equipment. The model was extended to Level 2. Residual risk was estimated at both Level 1 and Level 2, in terms of core damage frequency （CDF） and large early release frequency （LERF）, respectively.
文摘This paper presents main results of a comprehensive study performed by the European Clearinghouse on Operating Experience Feedback of NPP with the support of IRSN (Institut de S?reté Nucléaire et de Radioprotection) and GRS (Gesellschaft für Anlagen und Reaktorsi-cherheit mbH). Physical ageing mechanisms of structure, systems and components that eventually lead to ageing related systems and components failures at nuclear power plants were the main focus of this study. The analysis of ageing related events involved operating experience reported by NPP operators in France, Germany, USA and to the IAEA/NEA International Reporting system, on operating experience for the past 20 years (i.e. 1990-2009). A list of ageing related events was populated. Each ageing related event contained in the list was analyzed and results of analysis were summarized for each commodity group for which the ageing degradation appeared to be a dominant contributor or direct cause. The most common degradation mechanisms/ageing effects for each specific component/commodity group, their risk significance and consequences to the plant performance are described. This paper provides insights into ageing related operating experience as well as recommendations to deal with the physical ageing of nuclear power plant SSC important to safety.
文摘The conflict between safety and economics requirements is peculiar to the present nuclear power (NP). The main point of the conflict is that for traditional type reactors the increase of requirements to safety of nuclear power plants (NPP) worsens their economical characteristics. This is caused by large potential energy accumulated in reactor coolant. In the presented paper the opportunity and expediency of changeover to reactors with heavy liquid-metal coolants (HLMC) in future NP is grounded. First of all, this refers to lead-bismuth coolant (LBC) mastered in the process of operating nuclear submarines (NS) reactors. The reactor facilities (RFs) of that type cannot cause destruction of defense barriers and make possible deterministic elimination of severe accidents with catastrophic radioactivity release. So it will make possible to eliminate the highlighted conflict and reasons for existence of population’s radiophobia. Lead-bismuth fast reactor SVBR-100 with electric power of 100 MWe is the reactor facility of that type. The effect of accumulated in coolant potential energy on safety and economics is considered. Main specific features of SVBR-100 technology providing a high level of inherent self-protection and passive safety are presented.
基金Acknowledgements The authors gratefully acknowledge the financial support from the National Natural Science Foundation of China （Grant No. 51575480） and the Fundamental Research Funds for the Central Universities （Grant No. 2017FZA4012）.
文摘Polyethylene （PE） pipe, particularly high- density polyethylene （HDPE） pipe, has been successfully utilized to transport cooling water for both non-safety- and safety-related applications in nuclear power plant （NPP）. Though ASME Code Case N755, which is the first code case related to NPP HDPE pipe, requires a thorough nondestructive examination （NDE） of HDPE joints. However, no executable regulations presently exist because of the lack of a feasible NDE technique for HDPE pipe in NPP. This work presents a review of current developments in NDE technology for both HDPE pipe in NPP with a diameter of less than 400 mm and that of a larger size. For the former category, phased array ultrasonic technique is proven effective for inspecting typical defects in HDPE pipe, and is thus used in Chinese national standards GB/T 29460 and GB/T 29461. A defect- recognition technique is developed based on pattern recognition, and a safety assessment principle is summa- rized from the database of destructive testing. On the other hand, recent research and practical studies reveal that in current ultrasonic-inspection technology, the absence of effective ultrasonic inspection for large size was lack of consideration of the viscoelasticity effect of PE on acoustic wave propagation in current ultrasonic inspection technology. Furthermore, main technical problems were analyzed in the paper to achieve an effective ultrasonic test method in accordance to the safety and efficiency requirements of related regulations and standards. Finally, the development trend and challenges of NDE test technology for HDPE in NPP are discussed.
文摘Many developing countries need ecologically clean power sources (PS). The nuclear power plants are such sources. However, a great number of the developing countries do not possess developed large capacity power systems. Moreover, currently in the developing countries, there are no highly skilled personnel to provide construction and reliable and safe operation of the nuclear plants, which are complex and potentially hazardous systems. In some countries, the level of terroristic threat is extremely high. For that reason, there are specific requirements to the nuclear PSs intended for use in the developing countries. In the presented report, the specific requirements which must be met by the NPT proposed for use in developing countries are formulated, basic statements of the SVBR-100 concept are presented, design and principal scheme of the reactor fa-ility are described, major characteristics of SVBR-100 are summarized.
文摘This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behavior at shutdown reactor during maintenance. For the purpose of the analysis is selected a plant operating state with unsealed primary circuit by removing the MCP head. The reference nuclear power plant is Unit 6 at Kozloduy NPP (KNPP) site. RELAP5/ MOD3.2 computer code has been used to simulate the transient for WWER-1000/V320 NPP model. A model of WWER-1000 based on Unit 6 of KNPP has been developed for the RELAP5/MOD3.2 code at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS), Sofia. The plant modifications performed in frame of modernization program have been taken into account for the investigated conditions for the unsealed primary circuit. The most specific in this analysis compared to the analyses of NPP accidents at full power is the unavailability of some important safety systems. For the purpose of the present investigation two scenarios have been studied, involving a different number of safety systems with and without operator actions. The selected initiating event and scenarios are used in support of analytical validation of Emergency Operating Procedures (EOP) at low power and they are based on the suggestions of leading KNPP experts and are important in support of analytical validation of EOP at low power.
