R. A. P. Dwijayanto, Harun Ardiansyah, A. W. Harto
Thermal molten salt reactors can be designed in many configurations. This paper investigates the optimal geometry of a one fluid molten salt reactor (OFMSR) in a virtual one-and-half fluid configuration with a fixed fuel salt volume. Two primary configurations were studied, axial blanket (three models) and radial blanket (two models). Neutronic calculations were performed using MCNP6.2 and Serpent-2 reactor physics codes with ENDF/B-VII.0 continuous neutron library. The analysis comprises criticality calculation, temperature coefficient of reactivity (TCR), breeding ratio (BR), and kinetic parameters. The results imply a good agreement between MCNP and Serpent calculations. TCR values show a different pattern between axial and radial blanket configuration. Whilst the correlation between TCR and BR is inversely correlated in axial blanket, it is linear in radial blanket configuration. Overall, radial blanket configuration seemed to show better neutronic performance than axial blanket configuration, with comparably strong negative TCR and large BR.
{"title":"Verification and Geometry Optimization of a One Fluid Molten Salt Reactor (OFMSR) with Fixed Volume","authors":"R. A. P. Dwijayanto, Harun Ardiansyah, A. W. Harto","doi":"10.1115/1.4064465","DOIUrl":"https://doi.org/10.1115/1.4064465","url":null,"abstract":"\u0000 Thermal molten salt reactors can be designed in many configurations. This paper investigates the optimal geometry of a one fluid molten salt reactor (OFMSR) in a virtual one-and-half fluid configuration with a fixed fuel salt volume. Two primary configurations were studied, axial blanket (three models) and radial blanket (two models). Neutronic calculations were performed using MCNP6.2 and Serpent-2 reactor physics codes with ENDF/B-VII.0 continuous neutron library. The analysis comprises criticality calculation, temperature coefficient of reactivity (TCR), breeding ratio (BR), and kinetic parameters. The results imply a good agreement between MCNP and Serpent calculations. TCR values show a different pattern between axial and radial blanket configuration. Whilst the correlation between TCR and BR is inversely correlated in axial blanket, it is linear in radial blanket configuration. Overall, radial blanket configuration seemed to show better neutronic performance than axial blanket configuration, with comparably strong negative TCR and large BR.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"71 10","pages":""},"PeriodicalIF":0.4,"publicationDate":"2024-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139440686","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Molten salt reactors refers to a broad class of nuclear reactors that use a molten alkali-halide salt as the primary coolant fluid. This paper pertains to thermal spectrum liquid fuel molten fluoride salt reactors with graphite moderator (MSRs), where the molten salt also dissolves the actinide fuel. Xenon isotope 135, 135Xe, is a fission product that is produced during nuclear energy production and it acts as a neutron poison. Due to the circulating nature of the fuel salt in MSRs, there is a qualitative difference in the behavior of 135Xe in an MSR compared to a solid fueled reactor. Some of the 135Xe produced in fission may end up in the pore space of the graphite moderator used in a MSR. This paper examines the transfer and storage of 135Xe in MSR graphite. Prior publications are reviewed, the porosity of the MSR graphite is examined, governing equations are detailed, film layer production and destruction is discussed, the graphite / salt interface is explored, transport pathways are considered, transfer processes are exposited, the effect of charged species is examined, the solubility of noble gases in molten fluoride salts is examined, the mass diffusion coefficient in molten salts is explored, and the calculation of mass transfer coefficients is described.
{"title":"The Transfer of Xenon-135 to Molten Salt Reactor Graphite","authors":"Terry Price, Ondrej Chvala","doi":"10.1115/1.4064464","DOIUrl":"https://doi.org/10.1115/1.4064464","url":null,"abstract":"\u0000 Molten salt reactors refers to a broad class of nuclear reactors that use a molten alkali-halide salt as the primary coolant fluid. This paper pertains to thermal spectrum liquid fuel molten fluoride salt reactors with graphite moderator (MSRs), where the molten salt also dissolves the actinide fuel. Xenon isotope 135, 135Xe, is a fission product that is produced during nuclear energy production and it acts as a neutron poison. Due to the circulating nature of the fuel salt in MSRs, there is a qualitative difference in the behavior of 135Xe in an MSR compared to a solid fueled reactor. Some of the 135Xe produced in fission may end up in the pore space of the graphite moderator used in a MSR. This paper examines the transfer and storage of 135Xe in MSR graphite. Prior publications are reviewed, the porosity of the MSR graphite is examined, governing equations are detailed, film layer production and destruction is discussed, the graphite / salt interface is explored, transport pathways are considered, transfer processes are exposited, the effect of charged species is examined, the solubility of noble gases in molten fluoride salts is examined, the mass diffusion coefficient in molten salts is explored, and the calculation of mass transfer coefficients is described.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"68 15","pages":""},"PeriodicalIF":0.4,"publicationDate":"2024-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139440891","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Challenges with safeguarding molten salt reactor (MSR) designs have prompted the search for enhanced safeguards technologies and revised safeguards materials control & accountancy (MC&A) approaches. A molten salt sampling system is a subsystem being developed to help support facility MC&A in future MSRs by removing salt samples from the primary fuel and/or coolant salt loop of an MSR for chemical and isotopic analysis. To consider the safeguards implications of this molten salt sampling system early in the design process, we employed a safeguards by design approach during the development of a prototype molten salt sampling system. Specifically, we identified and tailored a checklist approach to systematically evaluate the design against recognized safeguards and security attributes. This technical brief describes the molten salt sampling system design and operational concept upon which we applied the safeguards by design methodology, conveys the methods we used to employ the safeguards by design approach on the molten salt sampling system design and discusses the preliminary results and design insights gained from this safeguards by design assessment.
{"title":"Technical Brief: Safeguardability Analysis of a Molten Salt Sampling System Design","authors":"M. Harkema, Steven Krahn, Paul Marotta","doi":"10.1115/1.4064343","DOIUrl":"https://doi.org/10.1115/1.4064343","url":null,"abstract":"Challenges with safeguarding molten salt reactor (MSR) designs have prompted the search for enhanced safeguards technologies and revised safeguards materials control & accountancy (MC&A) approaches. A molten salt sampling system is a subsystem being developed to help support facility MC&A in future MSRs by removing salt samples from the primary fuel and/or coolant salt loop of an MSR for chemical and isotopic analysis. To consider the safeguards implications of this molten salt sampling system early in the design process, we employed a safeguards by design approach during the development of a prototype molten salt sampling system. Specifically, we identified and tailored a checklist approach to systematically evaluate the design against recognized safeguards and security attributes. This technical brief describes the molten salt sampling system design and operational concept upon which we applied the safeguards by design methodology, conveys the methods we used to employ the safeguards by design approach on the molten salt sampling system design and discusses the preliminary results and design insights gained from this safeguards by design assessment.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"360 2","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-12-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139170134","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A reliable high-temperature molten salt pump is critical for the development of Fluoride-salt-cooled High-temperature Reactors (FHRs). By supporting the rotating journal, the suitable journal bearing can ensure that the high-temperature molten salt pump runs smoothly and efficiently in the high-temperature fluoride salt over a long period of time. However, many bearing candidates served well for only a short period and experienced several issues. Moreover, the molten salt pump journal misalignment or not is a key factor for the molten salt pump's long-term steady running. In the long-term operation, a misalignment in the journal bearing can result in vibrations and excessive wear on the bearing surface of the molten salt pump. The journal bearing dynamic characteristics is a meaningful sign to accurately assess the journal misalignment. Therefore, it is necessary to investigate the detailed journal bearing dynamic behavior under the high-temperature hydrodynamic fluoride salt lubrication conditions for FHR applications. This study's small amplitude vibration is superimposed on a steady-running journal bearing condition. A FORTRAN 90 program has been written for the journal bearing dynamic behavior analysis. The numerical results are validated with experimental data from the literature. The validated program was employed to predict the dynamic coefficients of high-temperature fluoride salt hydrodynamic lubricated journal bearing various Sommerfeld numbers. This study evaluating the journal bearing dynamic coefficients for molten salt pumps provides guidelines that are helpful for designing molten salt primary pumps.
{"title":"Molten Salt Pump Journal-Bearings Dynamic Characteristics Under Hydrodynamic Lubrication Conditions","authors":"Yuqi Liu, Minghui Chen, S. Che, Adam Burak","doi":"10.1115/1.4064336","DOIUrl":"https://doi.org/10.1115/1.4064336","url":null,"abstract":"A reliable high-temperature molten salt pump is critical for the development of Fluoride-salt-cooled High-temperature Reactors (FHRs). By supporting the rotating journal, the suitable journal bearing can ensure that the high-temperature molten salt pump runs smoothly and efficiently in the high-temperature fluoride salt over a long period of time. However, many bearing candidates served well for only a short period and experienced several issues. Moreover, the molten salt pump journal misalignment or not is a key factor for the molten salt pump's long-term steady running. In the long-term operation, a misalignment in the journal bearing can result in vibrations and excessive wear on the bearing surface of the molten salt pump. The journal bearing dynamic characteristics is a meaningful sign to accurately assess the journal misalignment. Therefore, it is necessary to investigate the detailed journal bearing dynamic behavior under the high-temperature hydrodynamic fluoride salt lubrication conditions for FHR applications. This study's small amplitude vibration is superimposed on a steady-running journal bearing condition. A FORTRAN 90 program has been written for the journal bearing dynamic behavior analysis. The numerical results are validated with experimental data from the literature. The validated program was employed to predict the dynamic coefficients of high-temperature fluoride salt hydrodynamic lubricated journal bearing various Sommerfeld numbers. This study evaluating the journal bearing dynamic coefficients for molten salt pumps provides guidelines that are helpful for designing molten salt primary pumps.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"24 5","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-12-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139168605","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
An improved heat flux partitioning model of pool boiling is proposed in this study to predict the material-conjugated pool boiling curve. The fundamental rationale behind the improved model is that the heat convection is only governed by far-field mechanisms while the heat quenching and evaporation are partially subjected to near-field material-dependent mechanisms. The quenching heat flux is derived dependently on thermal-effusivities of solid and liquid respectively based on the transient heat conduction analyses. The evaporative heat flux correlates the material-dependent bubble dynamics parameters including bubble departure frequency and nucleation site density together to yield a new analytical form and support the theoretical reflections of material-conjugated boiling behaviors. The proposed model can approximately capture the material-related impacts on boiling heat transfer coefficients and simulate pool boiling curves validated by the use of experimental results.
{"title":"An Improved Heat Flux Partitioning Model of Nucleate Boiling Under Saturated Pool Boiling Condition","authors":"Mingfu He, Minghui Chen","doi":"10.1115/1.4064337","DOIUrl":"https://doi.org/10.1115/1.4064337","url":null,"abstract":"An improved heat flux partitioning model of pool boiling is proposed in this study to predict the material-conjugated pool boiling curve. The fundamental rationale behind the improved model is that the heat convection is only governed by far-field mechanisms while the heat quenching and evaporation are partially subjected to near-field material-dependent mechanisms. The quenching heat flux is derived dependently on thermal-effusivities of solid and liquid respectively based on the transient heat conduction analyses. The evaporative heat flux correlates the material-dependent bubble dynamics parameters including bubble departure frequency and nucleation site density together to yield a new analytical form and support the theoretical reflections of material-conjugated boiling behaviors. The proposed model can approximately capture the material-related impacts on boiling heat transfer coefficients and simulate pool boiling curves validated by the use of experimental results.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"229 ","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-12-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139170326","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
With the development of micro-reactors, a Free-Piston Stirling Engine (FPSE) is a great candidate for the power conversion unit. Based on the advantages of the micro-reactor such as the compact design, long lasting, highly efficiency, and remote-control operation, an FPSE can provide almost the same as the requirements. In this paper, a 20-kW electric FPSE is proposed to support the development of the power conversion unit for microreactor application. The calculation method was done through MATLAB to analyze the design with all the significant losses in the engine. Through various designs and operating conditions for the engine, the proposed design has 21.4 percent efficiency with a total output power of 20.7 kW electric. With the testing through different parameters in the engine, the current design is well optimized to balance all the constraints which offer highly efficient, compact design, and reliability. Additionally, there is room for improvement during the design process, such as using the heat flux instead of a heat exchanger, robust foil for the regenerator, and simulation through 3D modeling to maximize the potential of the design. This study provides theoretical support for the design and analysis of the FPSE for micro-reactor applications.
{"title":"Design and Analysis of a Free-Piston Stirling Engine for Microreactor Applications","authors":"Phat Doan, Minghui Chen","doi":"10.1115/1.4064335","DOIUrl":"https://doi.org/10.1115/1.4064335","url":null,"abstract":"With the development of micro-reactors, a Free-Piston Stirling Engine (FPSE) is a great candidate for the power conversion unit. Based on the advantages of the micro-reactor such as the compact design, long lasting, highly efficiency, and remote-control operation, an FPSE can provide almost the same as the requirements. In this paper, a 20-kW electric FPSE is proposed to support the development of the power conversion unit for microreactor application. The calculation method was done through MATLAB to analyze the design with all the significant losses in the engine. Through various designs and operating conditions for the engine, the proposed design has 21.4 percent efficiency with a total output power of 20.7 kW electric. With the testing through different parameters in the engine, the current design is well optimized to balance all the constraints which offer highly efficient, compact design, and reliability. Additionally, there is room for improvement during the design process, such as using the heat flux instead of a heat exchanger, robust foil for the regenerator, and simulation through 3D modeling to maximize the potential of the design. This study provides theoretical support for the design and analysis of the FPSE for micro-reactor applications.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"258 19","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-12-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139170741","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Within a hybrid energy system, it is fundamental to have accurate and reliable computational tools to predict the plants; behaviour under different operating conditions; compared to other energy sources, analysis methods for nuclear systems must provide detailed information on reactor criticality and fuel evolution. Thanks to the advancements in computational hardware, using three-dimensional codes to obtain a local description of the reactor core has now become feasible both for deterministic codes and for Monte Carlo (MC) codes. Those computational methods must be compared with experimental measurements to assess their reliability. For this reason, the 3D MC code SERPENT is currently being validated for Light Water Reactor (LWR) fuel cycle simulations. This work will compare the isotopic concentrations measured in a Post Irradiation Experiment and the results of the MC routine, examining the Takahama-3 assembly test case. From literature reports, roughly 35 nuclide species have been measured at different axial locations by destructive analysis following several radiochemical techniques. A sensitivity analysis to evaluate the impact of design features on the results was carried out investigating the cross-section libraries, the simulation time discretisation and the imposition of an axial time-varying temperature. During the process, systematic sources of geometry-related errors were analysed as well. Overall, the model showed good agreement with the experimental data under an acceptable error threshold. The sensitivity studies also showed how the prediction capability could be increased up to +6%, adopting a realistic temperature mesh for the fuel instead of a uniform temperature approach.
在混合能源系统中，最重要的是要有准确可靠的计算工具来预测发电厂在不同运行条件下的行为；与其他能源相比，核系统的分析方法必须提供反应堆临界状态和燃料演变的详细信息。由于计算硬件的进步，无论是确定性代码还是蒙特卡罗（MC）代码，使用三维代码获得反应堆堆芯的局部描述现已变得可行。这些计算方法必须与实验测量结果进行比较，以评估其可靠性。因此，三维 MC 代码 SERPENT 目前正在进行轻水反应堆（LWR）燃料循环模拟验证。这项工作将比较在辐照后实验中测得的同位素浓度和 MC 程序的结果，并对高滨-3 号装配测试案例进行研究。根据文献报告，通过采用多种放射化学技术进行破坏性分析，在不同轴向位置测量了大约 35 种核素。对横截面库、模拟时间离散化和轴向时变温度进行了敏感性分析，以评估设计特征对结果的影响。在此过程中，还分析了几何相关误差的系统来源。总体而言，在可接受的误差阈值范围内，模型与实验数据显示出良好的一致性。灵敏度研究还表明，如果采用现实的燃料温度网格而不是均匀温度方法，预测能力最多可提高 6%。
{"title":"A Monte Carlo Fuel Assembly Model Validation Adopting Post Irradiation Experiment Dataset","authors":"Lorenzo Loi, A. Cammi, S. Lorenzi, C. Introini","doi":"10.1115/1.4064308","DOIUrl":"https://doi.org/10.1115/1.4064308","url":null,"abstract":"\u0000 Within a hybrid energy system, it is fundamental to have accurate and reliable computational tools to predict the plants; behaviour under different operating conditions; compared to other energy sources, analysis methods for nuclear systems must provide detailed information on reactor criticality and fuel evolution. Thanks to the advancements in computational hardware, using three-dimensional codes to obtain a local description of the reactor core has now become feasible both for deterministic codes and for Monte Carlo (MC) codes. Those computational methods must be compared with experimental measurements to assess their reliability. For this reason, the 3D MC code SERPENT is currently being validated for Light Water Reactor (LWR) fuel cycle simulations. This work will compare the isotopic concentrations measured in a Post Irradiation Experiment and the results of the MC routine, examining the Takahama-3 assembly test case. From literature reports, roughly 35 nuclide species have been measured at different axial locations by destructive analysis following several radiochemical techniques. A sensitivity analysis to evaluate the impact of design features on the results was carried out investigating the cross-section libraries, the simulation time discretisation and the imposition of an axial time-varying temperature. During the process, systematic sources of geometry-related errors were analysed as well. Overall, the model showed good agreement with the experimental data under an acceptable error threshold. The sensitivity studies also showed how the prediction capability could be increased up to +6%, adopting a realistic temperature mesh for the fuel instead of a uniform temperature approach.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":" 14","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138962035","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The dominant nuclear reactor technologies that comprise the current global operating fleet were developed and deployed over a relatively short, mid-twentieth century, period spanning the 1950s and 60s. Four of these technologies were deployed at fleet scales and commercially exported. The historical record indicates a remarkably consistent process of phased technology development that enabled the commercialization of designs that would define the global nuclear marketplace, beginning with research and development (R&D) and advancing through test reactors, small and large demonstration reactors, and first commercial-scale units. Following proof-of-principle R&D, historical commercialization lead times (from decision to construction of a demonstration reactor to first commercial launch) ranged from 12 to16 years for these four commercial technologies. Key factors contributing to successful commercialization included durable government support for early R&D and varying degrees of public-private partnering through commercial launch. This partnering included arrangements for technical support, siting, facility ownership, nuclear material provision, and cost sharing. The policy environment was characterized by unambiguous government support; stabile, effective and informed government program management and oversight; and flexibility in the public-private partnership arrangements to promote technology development and demonstration. Government advocacy was structured to support progressively increasing industry independence and self-sufficiency. This experience is documented and analyzed in this paper to provide salient lessons and example program elements for contemporary efforts to stimulate development and commercialization of a new generation of advanced nuclear technologies through collaboration and public-private partnerships.
{"title":"Public-Private Partnering in Nuclear Reactor Development - Historical Review and Implications for Today","authors":"Steven Krahn, Andrew Sowder","doi":"10.1115/1.4064233","DOIUrl":"https://doi.org/10.1115/1.4064233","url":null,"abstract":"\u0000 The dominant nuclear reactor technologies that comprise the current global operating fleet were developed and deployed over a relatively short, mid-twentieth century, period spanning the 1950s and 60s. Four of these technologies were deployed at fleet scales and commercially exported. The historical record indicates a remarkably consistent process of phased technology development that enabled the commercialization of designs that would define the global nuclear marketplace, beginning with research and development (R&D) and advancing through test reactors, small and large demonstration reactors, and first commercial-scale units. Following proof-of-principle R&D, historical commercialization lead times (from decision to construction of a demonstration reactor to first commercial launch) ranged from 12 to16 years for these four commercial technologies. Key factors contributing to successful commercialization included durable government support for early R&D and varying degrees of public-private partnering through commercial launch. This partnering included arrangements for technical support, siting, facility ownership, nuclear material provision, and cost sharing. The policy environment was characterized by unambiguous government support; stabile, effective and informed government program management and oversight; and flexibility in the public-private partnership arrangements to promote technology development and demonstration. Government advocacy was structured to support progressively increasing industry independence and self-sufficiency. This experience is documented and analyzed in this paper to provide salient lessons and example program elements for contemporary efforts to stimulate development and commercialization of a new generation of advanced nuclear technologies through collaboration and public-private partnerships.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"10 8","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-12-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138586112","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In a large-scale plant such as a Nuclear Power Plant (NPP), thousands of process values are measured for the purpose of monitoring the plant performance and the system health. It is difficult for plant operators to constantly monitor all of the process values. We present a data-driven method to comprehensively monitor a large number of process values and detect early signs of anomalies, including unknown events, with few false positives. In order to learn the complex changing internal state of a NPP and accurately predict the normal process values, we have developed a two-stage autoencoder (TSAE), a type of neural network, composed of a time window autoencoder and a deviation autoencoder. TSAE realizes to detect anomalous signals during the plant transient conditions by collecting time-series data and learning the nonlinear temporal correlation among them. In the actual plant, some process values which are physically uncorrelated with each other happen to behave similarly (pseudo-correlation). Learning the pseudo-correlation by the algorithm causes false positives because the predicted values of unrelated process values are incorrectly correlated. Therefore, Toshiba has proposed the model classification concept of separating the process values into two groups based on physical correlation and applied a model structure of TSAE. As a result, it becomes possible to learn only with the process values that are physically correlated and enhance the performance of prediction/detection. We assessed the improved TSAE with simulated process values of a NPP and showed excellent performances with few false positives.
{"title":"Development of an Ai-Based Predictive Anomaly Detection System to Nuclear Power Plant","authors":"Ryota Miyake, Shinya Tominaga, Yusuke Terakado, Naoyuki Takado, Toshio Aoki, Chikashi Miyamoto, Susumu Naito, Yasunori Taguchi, Yuichi Kato, Kota Nakata","doi":"10.1115/1.4064123","DOIUrl":"https://doi.org/10.1115/1.4064123","url":null,"abstract":"In a large-scale plant such as a Nuclear Power Plant (NPP), thousands of process values are measured for the purpose of monitoring the plant performance and the system health. It is difficult for plant operators to constantly monitor all of the process values. We present a data-driven method to comprehensively monitor a large number of process values and detect early signs of anomalies, including unknown events, with few false positives. In order to learn the complex changing internal state of a NPP and accurately predict the normal process values, we have developed a two-stage autoencoder (TSAE), a type of neural network, composed of a time window autoencoder and a deviation autoencoder. TSAE realizes to detect anomalous signals during the plant transient conditions by collecting time-series data and learning the nonlinear temporal correlation among them. In the actual plant, some process values which are physically uncorrelated with each other happen to behave similarly (pseudo-correlation). Learning the pseudo-correlation by the algorithm causes false positives because the predicted values of unrelated process values are incorrectly correlated. Therefore, Toshiba has proposed the model classification concept of separating the process values into two groups based on physical correlation and applied a model structure of TSAE. As a result, it becomes possible to learn only with the process values that are physically correlated and enhance the performance of prediction/detection. We assessed the improved TSAE with simulated process values of a NPP and showed excellent performances with few false positives.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"65 ","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139245096","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Nowadays, there is an increasing demand for new SMR reactors with a wide range of applications, often classified as a new generation IV. reactors. Unfortunately, there is no commercially operating nuclear reactor meeting the characteristics of Gen. IV reactors in its technical design and features. Gen. IV nuclear reactors are intensively developed worldwide, including the Czech Republic. At least two general Gen. IV thermal neutron reactor concepts use graphite as a moderator or reflector, as do many concepts of the very popular small modular reactors. To support research activities linked with the development of these reactors, an appropriate experimental environment and resources simulating conditions expected in Gen. IV reactors with graphite are needed. The calculated data confirm the results obtained during previous research. The experiment at LR-0 with a graphite reflector gives better results of neutron flux distribution in the reflector due to the extra graphite reflector layer and central graphite plugs. Besides, the core arrangement is included in a set of experiments supporting the research of reactor cores with graphite reflectors. The main reason for this article is to support the development of a new functional sample of neutron instrumentation for Gen. IV reactors.
{"title":"Special Experimental Environment for Gen. IV Reactors with Graphite Reflector","authors":"Eva Vilimová, T. Peltán, R. Škoda","doi":"10.1115/1.4064124","DOIUrl":"https://doi.org/10.1115/1.4064124","url":null,"abstract":"Nowadays, there is an increasing demand for new SMR reactors with a wide range of applications, often classified as a new generation IV. reactors. Unfortunately, there is no commercially operating nuclear reactor meeting the characteristics of Gen. IV reactors in its technical design and features. Gen. IV nuclear reactors are intensively developed worldwide, including the Czech Republic. At least two general Gen. IV thermal neutron reactor concepts use graphite as a moderator or reflector, as do many concepts of the very popular small modular reactors. To support research activities linked with the development of these reactors, an appropriate experimental environment and resources simulating conditions expected in Gen. IV reactors with graphite are needed. The calculated data confirm the results obtained during previous research. The experiment at LR-0 with a graphite reflector gives better results of neutron flux distribution in the reflector due to the extra graphite reflector layer and central graphite plugs. Besides, the core arrangement is included in a set of experiments supporting the research of reactor cores with graphite reflectors. The main reason for this article is to support the development of a new functional sample of neutron instrumentation for Gen. IV reactors.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"90 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139244868","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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