Le Mei, Xiaochun Zhang, Junbao Zhang, Changlei Shao, Jialei Zhu, Ran Huang, Chongzhi Wu
In order to realize the steel liner underwater repairing of the spent fuel pool of the third generation nuclear power plant, the laser welding process tests were carried out step by step in three environments: air, shallow water and simulating-repairing of the spent fuel pool floor(high-pressure condition). Through the process optimization, the high-quality forming of the underwater laser welding of duplex stainless steel was realized, and the underwater local dry laser welding process suitable for the spent fuel pool floor of nuclear power plant was developed. The results of nondestructive testing (including visual testing, liquid penetrant testing, ultrasonic testing and radiographic testing) of welding test pieces under three environments were qualified, and the test results of properties (including tensile, impact, bending, intergranular corrosion, ferrite content) meet the standard requirements. The underwater weld performance is similar to that in the air environment, and the weld quality meets the requirements of the spent fuel pool construction standard, laying a technical foundation for the application of the spent fuel pool underwater repairing.
{"title":"Study On the Step by Step Process and Performance of Laser Welding for the Spent Fuel Pool Floor","authors":"Le Mei, Xiaochun Zhang, Junbao Zhang, Changlei Shao, Jialei Zhu, Ran Huang, Chongzhi Wu","doi":"10.1115/1.4063008","DOIUrl":"https://doi.org/10.1115/1.4063008","url":null,"abstract":"\u0000 In order to realize the steel liner underwater repairing of the spent fuel pool of the third generation nuclear power plant, the laser welding process tests were carried out step by step in three environments: air, shallow water and simulating-repairing of the spent fuel pool floor(high-pressure condition). Through the process optimization, the high-quality forming of the underwater laser welding of duplex stainless steel was realized, and the underwater local dry laser welding process suitable for the spent fuel pool floor of nuclear power plant was developed. The results of nondestructive testing (including visual testing, liquid penetrant testing, ultrasonic testing and radiographic testing) of welding test pieces under three environments were qualified, and the test results of properties (including tensile, impact, bending, intergranular corrosion, ferrite content) meet the standard requirements. The underwater weld performance is similar to that in the air environment, and the weld quality meets the requirements of the spent fuel pool construction standard, laying a technical foundation for the application of the spent fuel pool underwater repairing.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"97 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82441959","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 Europe the design extension conditions (DEC) were introduced after the Fukushima Dai-ichi accident as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. The objective of the study is to determine available elapsed time before core uncovery and needed DEC safety features for total loss of all feedwater (TLOFW) in a two-loop pressurized water reactor. RELAP5/MOD3.3 computer code has been used for calculations. The initiating event for TLOFW are multiple failures in which, besides the loss of main feedwater also the auxiliary feedwater is lost. The scenarios without DEC safety features and the scenarios with DEC safety features assumed have been simulated. The results showed that after TLOFW event initiation it is very important to trip the reactor as soon as possible. In case of loss of offsite power the reactor coolant pumps stop and the reactor very quickly trips on low reactor coolant pump flow. When normal operation systems are assumed the reactor trip occurs on low-low steam generator narrow level few tens of seconds after accident initiation, resulting in less time available before core uncovery occurence. The results for TLOFW scenarios with normal operation systems and DEC safety featured assumed demonstrated that secondary side bleed and feed can prevent core uncovery in case when no operator actions are credited before 30 minute. When primary side bleed and feed is used, less time is available for operator actions.
{"title":"Total Loss of Feedwater Analysis of PWR Using RELAP5","authors":"A. Prošek","doi":"10.1115/1.4063009","DOIUrl":"https://doi.org/10.1115/1.4063009","url":null,"abstract":"\u0000 In Europe the design extension conditions (DEC) were introduced after the Fukushima Dai-ichi accident as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. The objective of the study is to determine available elapsed time before core uncovery and needed DEC safety features for total loss of all feedwater (TLOFW) in a two-loop pressurized water reactor. RELAP5/MOD3.3 computer code has been used for calculations. The initiating event for TLOFW are multiple failures in which, besides the loss of main feedwater also the auxiliary feedwater is lost. The scenarios without DEC safety features and the scenarios with DEC safety features assumed have been simulated.\u0000 The results showed that after TLOFW event initiation it is very important to trip the reactor as soon as possible. In case of loss of offsite power the reactor coolant pumps stop and the reactor very quickly trips on low reactor coolant pump flow. When normal operation systems are assumed the reactor trip occurs on low-low steam generator narrow level few tens of seconds after accident initiation, resulting in less time available before core uncovery occurence. The results for TLOFW scenarios with normal operation systems and DEC safety featured assumed demonstrated that secondary side bleed and feed can prevent core uncovery in case when no operator actions are credited before 30 minute. When primary side bleed and feed is used, less time is available for operator actions.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"301 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82871641","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 heavy water reactor concept Teplator is a pressure channel type reactor with independent systems for the primary coolant and the moderator. The present study analyses the low-pressure moderator cooling system of Teplator during full-power operation. The moderator is heated from neutron thermalization, gamma rays absorption, fission product decay and decay of activation products. Additionally, heat transfer from the coolant channels has to be taken in the analyses of the moderator cooling system. Preliminary thermal-hydraulic analyses of the cooling system are supplemented by CFD simulations of heat and fluid flow in the moderator's vessel, emphasizing flow-type regimes. Results from CFD simulations showed that the buoyancy-dominated flow (case MF-22) resulted in a higher thermal stratification and high moderator temperature close to the upper plate of the moderator vessel. The inertia-dominated flow regime MF-90 resulted in good mixing of the moderator and a low thermal stratification in the vessel. Finally, the mid-mass flow rate regime MF-45 was identified as a transitional region from a buoyancy-dominated to a mix-type regime.
{"title":"CFD Calculations of Moderator Heat and Fluid Flow of Small Modular Heavy Water Reactor","authors":"T. Kořínek, R. Škoda, M. Lovecký, O. Burian","doi":"10.1115/1.4063007","DOIUrl":"https://doi.org/10.1115/1.4063007","url":null,"abstract":"\u0000 The heavy water reactor concept Teplator is a pressure channel type reactor with independent systems for the primary coolant and the moderator. The present study analyses the low-pressure moderator cooling system of Teplator during full-power operation. The moderator is heated from neutron thermalization, gamma rays absorption, fission product decay and decay of activation products. Additionally, heat transfer from the coolant channels has to be taken in the analyses of the moderator cooling system. Preliminary thermal-hydraulic analyses of the cooling system are supplemented by CFD simulations of heat and fluid flow in the moderator's vessel, emphasizing flow-type regimes. Results from CFD simulations showed that the buoyancy-dominated flow (case MF-22) resulted in a higher thermal stratification and high moderator temperature close to the upper plate of the moderator vessel. The inertia-dominated flow regime MF-90 resulted in good mixing of the moderator and a low thermal stratification in the vessel. Finally, the mid-mass flow rate regime MF-45 was identified as a transitional region from a buoyancy-dominated to a mix-type regime.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"59 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81450350","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}
Rhodium is one of the scarcest, most valuable, and useful platinum group metals, a strategically important material relied on heavily by automotive and electronics industries. The limited finite natural sources of Rh and exponentially increasing demands on these supplies mean that new sources are being sought to stabilise supplies and prices. Spent nuclear fuel (SNF) contains a significant quantity of Rh, though methods to recover this are purely conceptual at this point, due to the differing chemistry between SNF reprocessing and the methods used to recycle natural Rh. During SNF reprocessing, Rh partitions between aqueous nitric acid streams, where its speciation is complex, and insoluble fission product waste streams. Various techniques have been investigated for Rh recovery during SNF reprocessing for over 50 years, including solvent extraction, ion exchange, precipitation, and electrochemical methods, with tuneable approaches such as impregnated composites and ionic liquids receiving the most attention recently, assisted by more the comprehensive understanding of Rh speciation in nitric acid developed recently. The quantitative recovery of Rh within the SNF reprocessing ecosystem has remained elusive thus far, and as such, this review discusses the recent developments within the field, and strategies that could be applied to maximise the recovery of Rh from SNF.
{"title":"A Review of Opportunities and Methods for Recovery of Rhodium from Spent Nuclear Fuel during Reprocessing","authors":"B. Hodgson, J. Turner, Alistair F. Holdsworth","doi":"10.3390/jne4030034","DOIUrl":"https://doi.org/10.3390/jne4030034","url":null,"abstract":"Rhodium is one of the scarcest, most valuable, and useful platinum group metals, a strategically important material relied on heavily by automotive and electronics industries. The limited finite natural sources of Rh and exponentially increasing demands on these supplies mean that new sources are being sought to stabilise supplies and prices. Spent nuclear fuel (SNF) contains a significant quantity of Rh, though methods to recover this are purely conceptual at this point, due to the differing chemistry between SNF reprocessing and the methods used to recycle natural Rh. During SNF reprocessing, Rh partitions between aqueous nitric acid streams, where its speciation is complex, and insoluble fission product waste streams. Various techniques have been investigated for Rh recovery during SNF reprocessing for over 50 years, including solvent extraction, ion exchange, precipitation, and electrochemical methods, with tuneable approaches such as impregnated composites and ionic liquids receiving the most attention recently, assisted by more the comprehensive understanding of Rh speciation in nitric acid developed recently. The quantitative recovery of Rh within the SNF reprocessing ecosystem has remained elusive thus far, and as such, this review discusses the recent developments within the field, and strategies that could be applied to maximise the recovery of Rh from SNF.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"11 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79205071","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}
Active special nuclear material (SNM) photoneutron interrogation research with Acoustically Tensioned Metastable Fluid Detector (ATMFD) sensor technology is discussed which provides evidence for enabling real time detection of special nuclear material (SNM) even when deployed under extreme 15,000 R h-1 (9 MeV endpoint) X-ray beams. Experiments to detect 3.2 kg DU are described with use of two designs of the economical E-ATMFD, viz., E-ATMFD.Ver.0 and E-ATMFD.Ver.1, respectively, at standoffs ranging from 0.1 m to 10 m - including with the E-ATMFD directly within the interrogating beam. Under similar conditions and with 100% photon rejection (i.e., 0 cpm with beam on, and w/o SNM), the E-ATMFD.Ver.1 design was shown capable of ~6x (600%) higher gain at ~10x lower drive powers over E-ATMFD.Ver.0 (with beam on and with SNM). The sensitivity gain rises to ~27x (i.e., 2,700%) with the E-ATMFD.Ver.1 operating at 0.99 W and a background count rate of ~1 cpm. The E-ATMFD.Ver.1 demonstrated 100% photon blindness (0 cpm) while operating at ~0.56 W drive power and placed directly within the beam under 15,000 R/h; including the SNM target led to a count rate of up to 50 cpm - revealing the E-ATMFD.Ver.1 as potentially field-capable for detecting U-based SNMs within seconds from photofission neutron signals, even when deployed directly within the interrogating photon beam.
讨论了利用声张力亚稳态流体检测器(ATMFD)传感器技术对特殊核材料(SNM)进行的有源光子中子探测研究，为在极端15000 R h-1 (9 MeV端点)x射线光束下实现特殊核材料(SNM)的实时探测提供了证据。描述了使用两种经济型E-ATMFD设计的3.2 kg DU检测实验，即E-ATMFD。0和E-ATMFD.Ver。1，分别在0.1米至10米的距离内-包括直接在询问光束内的E-ATMFD。在相似的条件下，100%光子抑制(即0 cpm，带光束，无SNM)， E-ATMFD.Ver。与E-ATMFD.Ver.0相比，1设计能够在低驱动功率约10倍的情况下获得约6倍(600%)的增益(有光束和SNM)。使用E-ATMFD.Ver，灵敏度增益提高到~27倍(即2,700%)。工作功率为0.99 W，背景计数速率为~1 cpm。E-ATMFD.Ver。在~0.56 W的驱动功率下，直接置于15000 R/h的光束中，证明了100%的光子盲性(0 cpm);包括SNM目标导致计数率高达50 cpm -揭示了E-ATMFD.Ver。1作为潜在的现场能力，可以在几秒钟内从光裂变中子信号中探测到基于u的SNMs，即使直接部署在询问光子束中。
{"title":"9Mv Linac Photo-Neutron Interrogation Of Uranium With Advanced Acoustically Tensioned Metastable Fluid Detectors","authors":"N. Boyle, S. Ozerov, C. Harabagiu, R. Taleyarkhan","doi":"10.1115/1.4062951","DOIUrl":"https://doi.org/10.1115/1.4062951","url":null,"abstract":"\u0000 Active special nuclear material (SNM) photoneutron interrogation research with Acoustically Tensioned Metastable Fluid Detector (ATMFD) sensor technology is discussed which provides evidence for enabling real time detection of special nuclear material (SNM) even when deployed under extreme 15,000 R h-1 (9 MeV endpoint) X-ray beams. Experiments to detect 3.2 kg DU are described with use of two designs of the economical E-ATMFD, viz., E-ATMFD.Ver.0 and E-ATMFD.Ver.1, respectively, at standoffs ranging from 0.1 m to 10 m - including with the E-ATMFD directly within the interrogating beam. Under similar conditions and with 100% photon rejection (i.e., 0 cpm with beam on, and w/o SNM), the E-ATMFD.Ver.1 design was shown capable of ~6x (600%) higher gain at ~10x lower drive powers over E-ATMFD.Ver.0 (with beam on and with SNM). The sensitivity gain rises to ~27x (i.e., 2,700%) with the E-ATMFD.Ver.1 operating at 0.99 W and a background count rate of ~1 cpm. The E-ATMFD.Ver.1 demonstrated 100% photon blindness (0 cpm) while operating at ~0.56 W drive power and placed directly within the beam under 15,000 R/h; including the SNM target led to a count rate of up to 50 cpm - revealing the E-ATMFD.Ver.1 as potentially field-capable for detecting U-based SNMs within seconds from photofission neutron signals, even when deployed directly within the interrogating photon beam.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"27 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88946558","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}
Seungyon Cho, Jeonghun Choi, J. Shin, Seung Jun Lee
Multi-abnormal events, referring to the simultaneous occurrence of multiple single abnormal events in a nuclear power plant, have not been subject to consideration because multi-abnormal events are extremely unlikely to occur and indeed have not yet occurred. Such events, though, would be more challenging to diagnose than general single abnormal events, exacerbating the human error issue. This study introduces an efficient abnormality diagnosis model that covers multi-abnormality diagnosis using a one-vs-rest classifier and compares it with other artificial intelligence models. The multi-abnormality attention diagnosis model deals with multi-label classification problems, for which two methods are proposed. First, a method to effectively cluster single and multi-abnormal events is introduced based on the predicted probability distribution of each abnormal event. Second, a one-vs-rest classifier with high accuracy is employed as an efficient way to obtain knowledge on which particular multi-abnormal events are the most difficult to diagnose and therefore require the most attention to improve the multi-label classification performance in terms of data usage. The developed multi-abnormality attention diagnosis model can reduce human errors of operators due to excessive information and limited time when unexpected multi-abnormal events occur by providing diagnosis results as part of an operator support system.
{"title":"Multi-Abnormality Attention Diagnosis Model Using One-vs-Rest Classifier in a Nuclear Power Plant","authors":"Seungyon Cho, Jeonghun Choi, J. Shin, Seung Jun Lee","doi":"10.3390/jne4030033","DOIUrl":"https://doi.org/10.3390/jne4030033","url":null,"abstract":"Multi-abnormal events, referring to the simultaneous occurrence of multiple single abnormal events in a nuclear power plant, have not been subject to consideration because multi-abnormal events are extremely unlikely to occur and indeed have not yet occurred. Such events, though, would be more challenging to diagnose than general single abnormal events, exacerbating the human error issue. This study introduces an efficient abnormality diagnosis model that covers multi-abnormality diagnosis using a one-vs-rest classifier and compares it with other artificial intelligence models. The multi-abnormality attention diagnosis model deals with multi-label classification problems, for which two methods are proposed. First, a method to effectively cluster single and multi-abnormal events is introduced based on the predicted probability distribution of each abnormal event. Second, a one-vs-rest classifier with high accuracy is employed as an efficient way to obtain knowledge on which particular multi-abnormal events are the most difficult to diagnose and therefore require the most attention to improve the multi-label classification performance in terms of data usage. The developed multi-abnormality attention diagnosis model can reduce human errors of operators due to excessive information and limited time when unexpected multi-abnormal events occur by providing diagnosis results as part of an operator support system.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"13 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90196767","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}
Small Modular Reactors (SMRs) are actively being considered for use in Canada. Some proposed SMRs can make use of solar salt as an intermediate coolant for a heat storage system. The development of thermalhydraulic simulation tools is one of the key capabilities needed to examine the performance of SMRs and license this class of reactors. This article summarizes the implementation of molten solar salt fluid properties into the ARIANT thermalhydraulic code and uses the code to simulate a high temperature gas-cooled SMR with helium and solar salt as its primary and secondary coolants during a pressurized loss of forced circulation (PLOFC) event. This work demonstrates the ability of ARIANT to simulate transient events in a two loop reactor system consisting of helium and solar salt as coolants and helps to establish ARIANT as a tool for SMR analysis.
{"title":"Implementation Of Solar Salt Properties Into Ariant And Simulation Of Pressurized Loss Of Forced Circulation In A High Temperature Gas-Cooled Small Modular Reactor","authors":"Jason Wu, T. Beuthe, Aleksandar Vasić","doi":"10.1115/1.4062917","DOIUrl":"https://doi.org/10.1115/1.4062917","url":null,"abstract":"\u0000 Small Modular Reactors (SMRs) are actively being considered for use in Canada. Some proposed SMRs can make use of solar salt as an intermediate coolant for a heat storage system. The development of thermalhydraulic simulation tools is one of the key capabilities needed to examine the performance of SMRs and license this class of reactors. This article summarizes the implementation of molten solar salt fluid properties into the ARIANT thermalhydraulic code and uses the code to simulate a high temperature gas-cooled SMR with helium and solar salt as its primary and secondary coolants during a pressurized loss of forced circulation (PLOFC) event. This work demonstrates the ability of ARIANT to simulate transient events in a two loop reactor system consisting of helium and solar salt as coolants and helps to establish ARIANT as a tool for SMR analysis.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"67 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73772349","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}
Jordan R. Stomps, Paul P. H. Wilson, K. Dayman, Michael J. Willis, James M. Ghawaly, Daniel E. Archer
The timely detection of special nuclear material (SNM) transfers between nuclear facilities is an important monitoring objective in nuclear nonproliferation. Persistent monitoring enabled by successful detection and characterization of radiological material movements could greatly enhance the nuclear nonproliferation mission in a range of applications. Supervised machine learning can be used to signal detections when material is present if a model is trained on sufficient volumes of labeled measurements. However, the nuclear monitoring data needed to train robust machine learning models can be costly to label since radiation spectra may require strict scrutiny for characterization. Therefore, this work investigates the application of semi-supervised learning to utilize both labeled and unlabeled data. As a demonstration experiment, radiation measurements from sodium iodide (NaI) detectors are provided by the Multi-Informatics for Nuclear Operating Scenarios (MINOS) venture at Oak Ridge National Laboratory (ORNL) as sample data. Anomalous measurements are identified using a method of statistical hypothesis testing. After background estimation, an energy-dependent spectroscopic analysis is used to characterize an anomaly based on its radiation signatures. In the absence of ground-truth information, a labeling heuristic provides data necessary for training and testing machine learning models. Supervised logistic regression serves as a baseline to compare three semi-supervised machine learning models: co-training, label propagation, and a convolutional neural network (CNN). In each case, the semi-supervised models outperform logistic regression, suggesting that unlabeled data can be valuable when training and demonstrating value in semi-supervised nonproliferation implementations.
{"title":"SNM Radiation Signature Classification Using Different Semi-Supervised Machine Learning Models","authors":"Jordan R. Stomps, Paul P. H. Wilson, K. Dayman, Michael J. Willis, James M. Ghawaly, Daniel E. Archer","doi":"10.3390/jne4030032","DOIUrl":"https://doi.org/10.3390/jne4030032","url":null,"abstract":"The timely detection of special nuclear material (SNM) transfers between nuclear facilities is an important monitoring objective in nuclear nonproliferation. Persistent monitoring enabled by successful detection and characterization of radiological material movements could greatly enhance the nuclear nonproliferation mission in a range of applications. Supervised machine learning can be used to signal detections when material is present if a model is trained on sufficient volumes of labeled measurements. However, the nuclear monitoring data needed to train robust machine learning models can be costly to label since radiation spectra may require strict scrutiny for characterization. Therefore, this work investigates the application of semi-supervised learning to utilize both labeled and unlabeled data. As a demonstration experiment, radiation measurements from sodium iodide (NaI) detectors are provided by the Multi-Informatics for Nuclear Operating Scenarios (MINOS) venture at Oak Ridge National Laboratory (ORNL) as sample data. Anomalous measurements are identified using a method of statistical hypothesis testing. After background estimation, an energy-dependent spectroscopic analysis is used to characterize an anomaly based on its radiation signatures. In the absence of ground-truth information, a labeling heuristic provides data necessary for training and testing machine learning models. Supervised logistic regression serves as a baseline to compare three semi-supervised machine learning models: co-training, label propagation, and a convolutional neural network (CNN). In each case, the semi-supervised models outperform logistic regression, suggesting that unlabeled data can be valuable when training and demonstrating value in semi-supervised nonproliferation implementations.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"11 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84410960","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}
Freeze plug is an important passive safety system used in the molten salt reactors (MSRs). It enables automatic drainage of the liquid fuel from the core to the storage tanks in an emergency to stop nuclear fission chain reaction without any operator's action and electric power supply. The opening time, that is the time taken for the freeze plug to open, is therefore of considerable importance to ensure passive safety of the MSRs. In our previous studies, systematic numerical simulations were carried out to understand how the fundamental design parameters such as the tube diameter and wall thickness of the freeze plug affected the opening time. In this work, a simple analytical model was developed for rough estimation of the opening time. It was shown that the opening time calculated by the present simple model was in fairly good agreement with that by the full simulation using the mass, momentum and energy conservation equations for the salt and the heat conduction equation within the wall material. The present simple model was hence shown to be useful particularly for the schematic design of the improved MSR freeze plugs.
{"title":"A Simple Analytical Model to Predict the Freeze Plug Opening Time in Molten Salt Reactors","authors":"M. Ilham, T. Okawa","doi":"10.1115/1.4062879","DOIUrl":"https://doi.org/10.1115/1.4062879","url":null,"abstract":"\u0000 Freeze plug is an important passive safety system used in the molten salt reactors (MSRs). It enables automatic drainage of the liquid fuel from the core to the storage tanks in an emergency to stop nuclear fission chain reaction without any operator's action and electric power supply. The opening time, that is the time taken for the freeze plug to open, is therefore of considerable importance to ensure passive safety of the MSRs. In our previous studies, systematic numerical simulations were carried out to understand how the fundamental design parameters such as the tube diameter and wall thickness of the freeze plug affected the opening time. In this work, a simple analytical model was developed for rough estimation of the opening time. It was shown that the opening time calculated by the present simple model was in fairly good agreement with that by the full simulation using the mass, momentum and energy conservation equations for the salt and the heat conduction equation within the wall material. The present simple model was hence shown to be useful particularly for the schematic design of the improved MSR freeze plugs.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"6 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-06-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89662526","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. Tamura, Yuki Hidaka, Haruhiko Ikeda, Norikazu Hamaura
With further applications of AI, IoT, and digital twin technology to plant operation and maintenance, it is becoming increasingly important to ensure data reliability. Data validation and reconciliation (DVR) represents one promising technique to ensure data reliability by minimizing the uncertainty of measurements based on statistics. DVR has been widely applied to nuclear power electrical generation plants in Europe and the United States in recent years. The most important input for DVR analysis is measurement uncertainty. In Japan, performance management of nuclear power plants is often done by measuring condensate flow rate. While the uncertainty of other flowmeters is handled by the JIS standard, the condensate flowmeter is specially calibrated every few cycles. This leads to reduction of effectiveness of DVR analysis due to variations in measurement uncertainty management. To overcome this issue, we propose an estimation method for measurement uncertainty by utilizing process data, an incidence matrix between sensors, and a reference instrument. The conventional method proposed in the previous study only treats the random error. The proposed method quantitatively estimates not only random error but also bias error by considering the uncertainty of the reference instrument. Using several benchmark problems, we found that the proposed method was applicable to various flow conditions, including physically fluctuating flow such as that observed in the feedwater flow in nuclear power plants. We anticipate that the proposed method will promote use of DVR analysis in nuclear power plants in Japan.
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