Roberta Ferri, Andrea Achilli, Cinzia Congiu, Stefano Marcianò, Stefano Gandolfi, Mattia Marengoni, Alberto Bersani, Alessandro Passerin D’Entreves
This paper summarises the results of an experimental campaign carried out at SIET on the ELSMOR facility built in 2022 to validate a decay heat removal system for the E-SMR. Based on the passive mechanisms of natural circulation, the system aims to dissipate the reactor decay heat to a water pool, using two heat exchangers: a plate-type one coupling the primary side to the secondary side, and a vertical tube one coupling the secondary side to the water pool. Such a system is considered to be the most effective passive system, capable of safely managing the SMR accident and accidental situations, and achieving long-term decay heat removal without the need for electricity or external inputs. A description of the primary and secondary loops of the plant is given, together with the installed instrumentation and data acquisition system. In addition, the paper summarises the tests performed in terms of test procedures, test type and associated objectives, test matrix, test results, achievements, and open issues.
{"title":"ELSMOR European Project: Experimental Results on an Innovative Decay Heat Removal System Based on a Plate-Type Heat Exchanger","authors":"Roberta Ferri, Andrea Achilli, Cinzia Congiu, Stefano Marcianò, Stefano Gandolfi, Mattia Marengoni, Alberto Bersani, Alessandro Passerin D’Entreves","doi":"10.1155/2023/6672504","DOIUrl":"https://doi.org/10.1155/2023/6672504","url":null,"abstract":"This paper summarises the results of an experimental campaign carried out at SIET on the ELSMOR facility built in 2022 to validate a decay heat removal system for the E-SMR. Based on the passive mechanisms of natural circulation, the system aims to dissipate the reactor decay heat to a water pool, using two heat exchangers: a plate-type one coupling the primary side to the secondary side, and a vertical tube one coupling the secondary side to the water pool. Such a system is considered to be the most effective passive system, capable of safely managing the SMR accident and accidental situations, and achieving long-term decay heat removal without the need for electricity or external inputs. A description of the primary and secondary loops of the plant is given, together with the installed instrumentation and data acquisition system. In addition, the paper summarises the tests performed in terms of test procedures, test type and associated objectives, test matrix, test results, achievements, and open issues.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":"16 1","pages":""},"PeriodicalIF":1.1,"publicationDate":"2023-12-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138632428","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
To investigate the seismic performance and isolation effect of a high-temperature gas-cooled reactor, a 1/20 scale model including a reactor, a spent-fuel plant, and a nuclear auxiliary plant was fabricated. In addition, 220 mm lead-rubber bearings were designed and produced for use in the shaking table test, which included both isolated and nonisolated conditions. Two historical earthquake records and three artificial earthquake motions were used to input the ground motion in the tests. The results demonstrated that the seismic performance of the plant was better and that the structure was in an elastic state, under a safe shutdown earthquake event. Isolation bearings were found to effectively reduce the dominate frequency of the structure. The acceleration amplification factor of the superstructure was found to be less than 1. The isolation test results showed that the peak of the floor response spectrum at the pressure vessel support was less than 0.1 g. In the nonisolation test, the peak of the floor response spectrum was greater than 1 g. In the isolation test, the relative displacement of the structure was less than 1.1 mm, which was relatively small. The structure maintained a good isolation performance and exhibited improved safety under extreme ground motion.
{"title":"Shaking Table Testing of a Scaled Nuclear Power Plant Structure with Base Isolation","authors":"Linlin Song, Xueming Zhang, Mingyang Wei, Yunlun Sun, Shicai Chen, Yan Chen","doi":"10.1155/2023/2536474","DOIUrl":"https://doi.org/10.1155/2023/2536474","url":null,"abstract":"To investigate the seismic performance and isolation effect of a high-temperature gas-cooled reactor, a 1/20 scale model including a reactor, a spent-fuel plant, and a nuclear auxiliary plant was fabricated. In addition, 220 mm lead-rubber bearings were designed and produced for use in the shaking table test, which included both isolated and nonisolated conditions. Two historical earthquake records and three artificial earthquake motions were used to input the ground motion in the tests. The results demonstrated that the seismic performance of the plant was better and that the structure was in an elastic state, under a safe shutdown earthquake event. Isolation bearings were found to effectively reduce the dominate frequency of the structure. The acceleration amplification factor of the superstructure was found to be less than 1. The isolation test results showed that the peak of the floor response spectrum at the pressure vessel support was less than 0.1 g. In the nonisolation test, the peak of the floor response spectrum was greater than 1 g. In the isolation test, the relative displacement of the structure was less than 1.1 mm, which was relatively small. The structure maintained a good isolation performance and exhibited improved safety under extreme ground motion.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":"33 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135808690","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Waad Saleh, Dalibor Kojecky, Edyta Agata Macieja, Juyoul Kim
This paper introduces the utilization of the International Atomic Energy Agency’s toolkit for reactor technology assessment (RTA) application to deploy small modular reactors (SMRs) in the Czech Republic, Egypt, and Poland. The increasing demand for clean energy has led to the prominence of small modular reactors (SMRs) in addressing global energy challenges. The successful integration of SMRs into national energy systems necessitates comprehensive evaluations that take into account each country’s specific characteristics and energy requirements. RTA application represents significant progress towards innovative nuclear solutions, advancing a cleaner and more resilient energy future plan. The aim of this study is assessing the feasibility and advantages of SMR implementation in these countries, focusing on energy security, emission reduction, and long-term sustainability. Various SMR technologies, including NuScale, SMART, HTR-PM, BWRX-300, SMR-160, and RITM-200, are comparatively analyzed based on safety, scalability, efficiency, and economic viability. The findings reveal that BWRX-300 suits the needs of the Czech Republic and Poland, while RITM-200 is the optimal choice for Egypt. Moreover, NuScale also stands as a strong alternative for all three countries. This article emphasizes the importance of informed discussions and evidence-based decisions, promoting sustainable energy development and global advancements in nuclear technology. By utilizing SMRs, the Czech Republic, Egypt, and Poland can enhance energy security, reduce emissions, and meet rising energy needs sustainably.
{"title":"Advancing Small Modular Reactor Technology Assessment in the Czech Republic, Egypt, and Poland","authors":"Waad Saleh, Dalibor Kojecky, Edyta Agata Macieja, Juyoul Kim","doi":"10.1155/2023/7002980","DOIUrl":"https://doi.org/10.1155/2023/7002980","url":null,"abstract":"This paper introduces the utilization of the International Atomic Energy Agency’s toolkit for reactor technology assessment (RTA) application to deploy small modular reactors (SMRs) in the Czech Republic, Egypt, and Poland. The increasing demand for clean energy has led to the prominence of small modular reactors (SMRs) in addressing global energy challenges. The successful integration of SMRs into national energy systems necessitates comprehensive evaluations that take into account each country’s specific characteristics and energy requirements. RTA application represents significant progress towards innovative nuclear solutions, advancing a cleaner and more resilient energy future plan. The aim of this study is assessing the feasibility and advantages of SMR implementation in these countries, focusing on energy security, emission reduction, and long-term sustainability. Various SMR technologies, including NuScale, SMART, HTR-PM, BWRX-300, SMR-160, and RITM-200, are comparatively analyzed based on safety, scalability, efficiency, and economic viability. The findings reveal that BWRX-300 suits the needs of the Czech Republic and Poland, while RITM-200 is the optimal choice for Egypt. Moreover, NuScale also stands as a strong alternative for all three countries. This article emphasizes the importance of informed discussions and evidence-based decisions, promoting sustainable energy development and global advancements in nuclear technology. By utilizing SMRs, the Czech Republic, Egypt, and Poland can enhance energy security, reduce emissions, and meet rising energy needs sustainably.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":"55 2","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135512489","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In the failure analysis and safety assessment of dissimilar metal welded joints, the mechanical heterogeneity of local regions is usually ignored and limited sampling locations are selected. The mechanical behavior of the crack tip region is the main variables affecting the environmentally assisted cracking behavior, and it is crucial for understanding the impact of mechanical heterogeneity on the local stress-strain state at the crack tip in welded joints. In this study, the effect of mechanical heterogeneity on the local mechanical behavior at the crack tip and on the stress-strain condition at the crack tip front for different crack sizes was investigated through finite-element simulations based on user-defined material subroutines. The local mechanical behavior of an interface region and crack propagation direction with mechanical heterogeneity and a series of initial crack locations were analyzed. The results show that mechanical heterogeneity has a significant effect on the mechanical condition and growth path of cracks at different sampling locations. The interaction between the mechanical heterogeneity around the crack and the crack depth determines the stress and plastic strain in front of the crack tip, which causes a substantial change in the crack growth path. The interface cracks have high stress and plastic strain; thus, the interface is often the weak position where damage occurs. To guarantee a reliable integrity assessment of cracks in mechanically heterogeneous interface regions, local mechanical properties related to crack locations should be determined and utilized.
{"title":"Effect of Inhomogeneous Mechanical Properties on the Stress-Strain Field at the Crack Tip and Crack Growth Direction in Dissimilar Metal Welded Joints","authors":"Shuang Wang, Hongkui Zhang, Zhe Ju, Bing Li, Fandong Chen, Fei Han","doi":"10.1155/2023/6408667","DOIUrl":"https://doi.org/10.1155/2023/6408667","url":null,"abstract":"In the failure analysis and safety assessment of dissimilar metal welded joints, the mechanical heterogeneity of local regions is usually ignored and limited sampling locations are selected. The mechanical behavior of the crack tip region is the main variables affecting the environmentally assisted cracking behavior, and it is crucial for understanding the impact of mechanical heterogeneity on the local stress-strain state at the crack tip in welded joints. In this study, the effect of mechanical heterogeneity on the local mechanical behavior at the crack tip and on the stress-strain condition at the crack tip front for different crack sizes was investigated through finite-element simulations based on user-defined material subroutines. The local mechanical behavior of an interface region and crack propagation direction with mechanical heterogeneity and a series of initial crack locations were analyzed. The results show that mechanical heterogeneity has a significant effect on the mechanical condition and growth path of cracks at different sampling locations. The interaction between the mechanical heterogeneity around the crack and the crack depth determines the stress and plastic strain in front of the crack tip, which causes a substantial change in the crack growth path. The interface cracks have high stress and plastic strain; thus, the interface is often the weak position where damage occurs. To guarantee a reliable integrity assessment of cracks in mechanically heterogeneous interface regions, local mechanical properties related to crack locations should be determined and utilized.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":"1 1","pages":""},"PeriodicalIF":1.1,"publicationDate":"2023-08-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41450916","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
V. Vo, V. Nguyen, Van-Diep Le, Q. Phan, L. Phan, Huy-Bach Nguyen, Nhi-Dien Nguyen
This paper introduces a controller unit for reactivity monitoring and automatic power control that was designed and constructed for the 500 kW Dalat Nuclear Research Reactor (DNRR). For power control and reactivity calculations, frequency signals from neutron measurement channels of starting and working ranges of the reactor are used. Two abovementioned independent functions were combined in an Artix-7 FPGA board for determining reactivity values by solving the point reactor kinetics equations with six delayed neutron groups and for stabilizing the reactor power at preset levels by determining the unbalance voltage signal to control the automatic control rod. With real-time calculations, the newly developed controller can monitor the reactor reactivity and control the reactor power online. The developed controller unit’s reactivity measuring and power stabilizing capabilities were assessed using the DNRR in normal operation and assumed emergency conditions and compared with those of the preexisting imported BNO-102R1 module of the DNRR control and protection system, known as ASUZ-14R. The results of the experiments show that the produced FPGA-based unit and the BNO-102R1 unit have the same technical characteristics and features, with the disparities being less than 5% and 1%, respectively, in reactivity measurement and power stabilization. The experimental data of reactivity measurements by the FPGA-based unit and the calculation results were also compared and found that the relative deviations between those are also less than 10%. The developed controller unit is capable of carrying out a variety of training and operational experiments on the DNRR.
{"title":"A Digital Controller for Reactivity Monitoring and Power Control","authors":"V. Vo, V. Nguyen, Van-Diep Le, Q. Phan, L. Phan, Huy-Bach Nguyen, Nhi-Dien Nguyen","doi":"10.1155/2023/2839654","DOIUrl":"https://doi.org/10.1155/2023/2839654","url":null,"abstract":"This paper introduces a controller unit for reactivity monitoring and automatic power control that was designed and constructed for the 500 kW Dalat Nuclear Research Reactor (DNRR). For power control and reactivity calculations, frequency signals from neutron measurement channels of starting and working ranges of the reactor are used. Two abovementioned independent functions were combined in an Artix-7 FPGA board for determining reactivity values by solving the point reactor kinetics equations with six delayed neutron groups and for stabilizing the reactor power at preset levels by determining the unbalance voltage signal to control the automatic control rod. With real-time calculations, the newly developed controller can monitor the reactor reactivity and control the reactor power online. The developed controller unit’s reactivity measuring and power stabilizing capabilities were assessed using the DNRR in normal operation and assumed emergency conditions and compared with those of the preexisting imported BNO-102R1 module of the DNRR control and protection system, known as ASUZ-14R. The results of the experiments show that the produced FPGA-based unit and the BNO-102R1 unit have the same technical characteristics and features, with the disparities being less than 5% and 1%, respectively, in reactivity measurement and power stabilization. The experimental data of reactivity measurements by the FPGA-based unit and the calculation results were also compared and found that the relative deviations between those are also less than 10%. The developed controller unit is capable of carrying out a variety of training and operational experiments on the DNRR.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2023-08-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45042450","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The neutronics and thermal-hydraulics (N/TH) coupling behavior analysis is a key issue for nuclear power plant design and safety analysis. Due to the high-dimensional partial differential equations (PDEs) derived from the N/TH system, it is usually time consuming to solve such a large-scale nonlinear equation by the traditional numerical solution method of PDEs. To solve this problem, this work develops a reduced order model based on the proper orthogonal decomposition (POD) and artificial neural networks (ANNs) to simulate the N/TH coupling system. In detail, the POD method is used to extract the POD modes and corresponding coefficients from a set of full-order model results under different boundary conditions. Then, the backpropagation neural network (BPNN) is utilized to map the relationship between the boundary conditions and POD coefficients. Therefore, the physical fields under the new boundary conditions could be calculated by the predicated POD coefficients from ANN and POD modes from snapshot. In order to assess the performance of an ANN-POD-based reduced order method, a simplified pressurized water reactor model under different inlet coolant temperatures and inlet coolant velocities is utilized. The results show that the new reduced order model can accurately predict the distribution of the physical fields, as well as the effective multiplication factor in the N/TH coupling nuclear system, whose relative errors are within 1%.
{"title":"A Reduced Order Model Based on ANN-POD Algorithm for Steady-State Neutronics and Thermal-Hydraulics Coupling Problem","authors":"Hanxing Liu, Han Zhang","doi":"10.1155/2023/9385756","DOIUrl":"https://doi.org/10.1155/2023/9385756","url":null,"abstract":"The neutronics and thermal-hydraulics (N/TH) coupling behavior analysis is a key issue for nuclear power plant design and safety analysis. Due to the high-dimensional partial differential equations (PDEs) derived from the N/TH system, it is usually time consuming to solve such a large-scale nonlinear equation by the traditional numerical solution method of PDEs. To solve this problem, this work develops a reduced order model based on the proper orthogonal decomposition (POD) and artificial neural networks (ANNs) to simulate the N/TH coupling system. In detail, the POD method is used to extract the POD modes and corresponding coefficients from a set of full-order model results under different boundary conditions. Then, the backpropagation neural network (BPNN) is utilized to map the relationship between the boundary conditions and POD coefficients. Therefore, the physical fields under the new boundary conditions could be calculated by the predicated POD coefficients from ANN and POD modes from snapshot. In order to assess the performance of an ANN-POD-based reduced order method, a simplified pressurized water reactor model under different inlet coolant temperatures and inlet coolant velocities is utilized. The results show that the new reduced order model can accurately predict the distribution of the physical fields, as well as the effective multiplication factor in the N/TH coupling nuclear system, whose relative errors are within 1%.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2023-07-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46151214","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In this paper, a nanosecond voltage comparator with PECL logic for a photon-counting radiation imaging system is presented. To realize a high-speed comparison of four gamma detector channels in a limited board space, quad comparators MAX9602 with PECL logic are chosen. Each of the four channels is coupled with a PECL to CMOS converter ICS508, which exports CMOS logic data for later use in an FPGA. Simulated findings for cobalt-60 with intensities ranging from 30 Ci to 300 Ci show little count loss caused by using a comparator and indicate ideal propagation delays at all source intensities. While in the laboratory test using a PCB-level system, signals with pulse width less than 3 ns might be dropped, and dispersion of propagation delay occurs. Despite these, the performance is still satisfactory and can meet the requirements of practical applications, as demonstrated by an improved result of 0.9% in the contrast indicator model. Further studies to optimize the circuit design can be conducted to gain improvement.
{"title":"Design of a Nanosecond Voltage Comparator with PECL Logic for a Photon-Counting Radiation Imaging System Application","authors":"Huaxia Zhang, Yuewen Sun, Zijia Chen, Zhifang Wu","doi":"10.1155/2023/6810882","DOIUrl":"https://doi.org/10.1155/2023/6810882","url":null,"abstract":"In this paper, a nanosecond voltage comparator with PECL logic for a photon-counting radiation imaging system is presented. To realize a high-speed comparison of four gamma detector channels in a limited board space, quad comparators MAX9602 with PECL logic are chosen. Each of the four channels is coupled with a PECL to CMOS converter ICS508, which exports CMOS logic data for later use in an FPGA. Simulated findings for cobalt-60 with intensities ranging from 30 Ci to 300 Ci show little count loss caused by using a comparator and indicate ideal propagation delays at all source intensities. While in the laboratory test using a PCB-level system, signals with pulse width less than 3 ns might be dropped, and dispersion of propagation delay occurs. Despite these, the performance is still satisfactory and can meet the requirements of practical applications, as demonstrated by an improved result of 0.9% in the contrast indicator model. Further studies to optimize the circuit design can be conducted to gain improvement.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2023-07-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48927956","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
When the feedwater valve at the outage loop of the floating nuclear power plant leaks, thermal stratification occurs in the steam generator. It causes lower water temperature in the outage loop. The extent of hazard of this phenomenon cannot be directly determined by the existing measurement parameters, which poses a threat to the operational safety of the reactor. Therefore, this study adopts two routes: data-driven combined with safety analysis system (DSAS) and mechanism model-driven combined with safety analysis system (MSAS), to propose the prediction methods for the minimum temperature of the outage loop and the maximum power caused by the low-temperature coolant. Then, the actual data are used to verify these methods and the prediction results under different initial conditions are analyzed. The results show that both the DSAS method and the MSAS method can predict the minimum temperature of the steam generator in the outage loop and the maximum power when the outage loop is put into operation, but the DSAS method has better performance under certain conditions. These methods can provide guidance to the operators to avoid reactivity insertion accident.
{"title":"Methods for Predicting the Minimum Temperature of the Outage Loop and the Maximum Power Caused by the Low-Temperature Coolant","authors":"Xinxin Liu, Lei Yu, J. Hao, Xiao-long Wang","doi":"10.1155/2023/4763033","DOIUrl":"https://doi.org/10.1155/2023/4763033","url":null,"abstract":"When the feedwater valve at the outage loop of the floating nuclear power plant leaks, thermal stratification occurs in the steam generator. It causes lower water temperature in the outage loop. The extent of hazard of this phenomenon cannot be directly determined by the existing measurement parameters, which poses a threat to the operational safety of the reactor. Therefore, this study adopts two routes: data-driven combined with safety analysis system (DSAS) and mechanism model-driven combined with safety analysis system (MSAS), to propose the prediction methods for the minimum temperature of the outage loop and the maximum power caused by the low-temperature coolant. Then, the actual data are used to verify these methods and the prediction results under different initial conditions are analyzed. The results show that both the DSAS method and the MSAS method can predict the minimum temperature of the steam generator in the outage loop and the maximum power when the outage loop is put into operation, but the DSAS method has better performance under certain conditions. These methods can provide guidance to the operators to avoid reactivity insertion accident.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2023-06-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47581952","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Probabilistic risk assessment (PRA) is an effective methodology that could be used to improve the safety of nuclear power plants in a reasonable manner. Dynamic PRA, as an advanced PRA, allows for more realistic and detailed analyses by handling time-dependent information. However, the applications of this method to practical problems are limited because it remains in the research and development stage. This study aimed to investigate the possibility of utilizing dynamic PRA in risk-informeddecision-making. Specifically, the author performed an accident sequence precursor (ASP) analysis on the failure of emergency diesel generators that occurred at Unit 1 of the Tomari Nuclear Power Plant in Japan using dynamic PRA. The results were evaluated by comparison with the results of simplified classical PRA. The findings indicated that dynamic PRA may estimate lower risks compared with those obtained from classical PRA by reasonable modeling of alternating current power recovery. The author also showed that dynamic PRA can provide detailed information that cannot be obtained with classical PRA, such as uncertainty distribution of core damage timing and importance measure considering the system failure timing.
{"title":"Accident Sequence Precursor Analysis of an Incident in a Japanese Nuclear Power Plant Based on Dynamic Probabilistic Risk Assessment","authors":"K. Kubo","doi":"10.1155/2023/7402217","DOIUrl":"https://doi.org/10.1155/2023/7402217","url":null,"abstract":"Probabilistic risk assessment (PRA) is an effective methodology that could be used to improve the safety of nuclear power plants in a reasonable manner. Dynamic PRA, as an advanced PRA, allows for more realistic and detailed analyses by handling time-dependent information. However, the applications of this method to practical problems are limited because it remains in the research and development stage. This study aimed to investigate the possibility of utilizing dynamic PRA in risk-informeddecision-making. Specifically, the author performed an accident sequence precursor (ASP) analysis on the failure of emergency diesel generators that occurred at Unit 1 of the Tomari Nuclear Power Plant in Japan using dynamic PRA. The results were evaluated by comparison with the results of simplified classical PRA. The findings indicated that dynamic PRA may estimate lower risks compared with those obtained from classical PRA by reasonable modeling of alternating current power recovery. The author also showed that dynamic PRA can provide detailed information that cannot be obtained with classical PRA, such as uncertainty distribution of core damage timing and importance measure considering the system failure timing.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2023-06-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41630974","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Establishing a dynamic model that accurately describes a realistic pressurized water reactor (PWR) fuel assembly is crucial to precisely evaluate the mechanical properties of the fuel assembly in seismic or loss of coolant accidents (LOCAs). The pluck test combined with the logarithmic decrement method has been widely applied in previous studies to extract fundamental modal parameters to calibrate dynamic models. However, most previous investigations focused on the first cycle of free vibration, which is strongly affected by stiction, baseline shift, drop conditions, and high-order mode interference, leading to inaccurate results. Moreover, these traditional methods cannot be used to extract high-order modal parameters. In this work, a novel experimental method for identifying the nonlinear modal parameters of a PWR fuel assembly is proposed. First, two algorithms are adopted to decompose the free vibration. Second, the local linearized modal parameters are extracted by a single-degree-of-freedom fitting method with a sliding window. Finally, these local linearized modal parameters are summed to obtain the nonlinear relationships between the modal parameters and amplitude. The new method makes more effective use of experimental data, obtains more accurate modal parameters than the logarithmic decrement method, and is capable of extracting high-order modal parameters. In the end, the test results are fitted by a fractional polynomial, which is of great value for numerical simulations.
{"title":"A New Experimental Method for the Nonlinear Modal Parameter Identification of a Pressurized Water Reactor Fuel Assembly","authors":"Chen Yang, Yan Guo, Xiao Hu, Yanhong Zhang","doi":"10.1155/2023/8892196","DOIUrl":"https://doi.org/10.1155/2023/8892196","url":null,"abstract":"Establishing a dynamic model that accurately describes a realistic pressurized water reactor (PWR) fuel assembly is crucial to precisely evaluate the mechanical properties of the fuel assembly in seismic or loss of coolant accidents (LOCAs). The pluck test combined with the logarithmic decrement method has been widely applied in previous studies to extract fundamental modal parameters to calibrate dynamic models. However, most previous investigations focused on the first cycle of free vibration, which is strongly affected by stiction, baseline shift, drop conditions, and high-order mode interference, leading to inaccurate results. Moreover, these traditional methods cannot be used to extract high-order modal parameters. In this work, a novel experimental method for identifying the nonlinear modal parameters of a PWR fuel assembly is proposed. First, two algorithms are adopted to decompose the free vibration. Second, the local linearized modal parameters are extracted by a single-degree-of-freedom fitting method with a sliding window. Finally, these local linearized modal parameters are summed to obtain the nonlinear relationships between the modal parameters and amplitude. The new method makes more effective use of experimental data, obtains more accurate modal parameters than the logarithmic decrement method, and is capable of extracting high-order modal parameters. In the end, the test results are fitted by a fractional polynomial, which is of great value for numerical simulations.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2023-05-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45553973","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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