Luca Berti, A. Pesetti, M. Raucci, Guglielmo Giambartolomei, D. Aquaro
At the Department of Civil and Industrial Engineering (DICI) of the University of Pisa, an experimental research program, funded by ITER Organization, concerning steam direct condensation in a flux containing also non-condensable gas and dust, was carried out. This mixture of fluids and dust is injected into the ITER Pressure Suppression Tanks during a Loss of Coolant Accident in the Vacuum Vessel. The aim of the research program is to determine the steam condensation efficiency in such conditions. Experimental tests were performed injecting this mixture in a tank partially filled with water. Alumina was used to simulate the actual dust present in the ITER Vacuum Vessel. Mass flow rates, temperature and pressure of the different fluids involved were recorded during the tests. The steam condensation into the subcooled water pool at a temperature ranging between 20 and 100°C was investigated to determine the condensation regimes occurring during the mixture injection. The values of the fraction of the energy absorbed by water, dust and metallic structures, of the heat losses and of the average heat transfer coefficient were determined considering pure steam, steam-dust and steam-air-dust injection. The average heat transfer coefficient, determined calculating the steam jet surfaces by means of image elaboration, was compared with empirical correlations.
比萨大学（University of Pisa）土木与工业工程系（DICI）在热核实验堆组织的资助下开展了一项实验研究计划，研究内容是蒸汽在同时含有非凝结气体和粉尘的通量中的直接凝结。在真空容器发生冷却剂损失事故时，这种流体和灰尘混合物会被注入热核实验堆压力抑制罐。研究计划的目的是确定这种情况下的蒸汽冷凝效率。实验测试将这种混合物注入部分注满水的水箱中。氧化铝被用来模拟热核实验堆真空容器中实际存在的灰尘。试验期间记录了不同流体的质量流量、温度和压力。 研究了蒸汽在 20 至 100 摄氏度之间的过冷水池中的冷凝情况，以确定混合物注入过程中出现的冷凝状态。考虑到纯蒸汽、蒸汽-粉尘和蒸汽-空气-粉尘喷射，确定了水、粉尘和金属结构吸收的能量比例值、热损失和平均传热系数。通过图像处理计算蒸汽喷射表面得出的平均传热系数与经验相关系数进行了比较。
{"title":"Influence of Non-Condensable Gas-Dust Mixture On Direct Contact Condensation of Steam At Atmospheric Pressure","authors":"Luca Berti, A. Pesetti, M. Raucci, Guglielmo Giambartolomei, D. Aquaro","doi":"10.1115/1.4064066","DOIUrl":"https://doi.org/10.1115/1.4064066","url":null,"abstract":"At the Department of Civil and Industrial Engineering (DICI) of the University of Pisa, an experimental research program, funded by ITER Organization, concerning steam direct condensation in a flux containing also non-condensable gas and dust, was carried out. This mixture of fluids and dust is injected into the ITER Pressure Suppression Tanks during a Loss of Coolant Accident in the Vacuum Vessel. The aim of the research program is to determine the steam condensation efficiency in such conditions. Experimental tests were performed injecting this mixture in a tank partially filled with water. Alumina was used to simulate the actual dust present in the ITER Vacuum Vessel. Mass flow rates, temperature and pressure of the different fluids involved were recorded during the tests. The steam condensation into the subcooled water pool at a temperature ranging between 20 and 100°C was investigated to determine the condensation regimes occurring during the mixture injection. The values of the fraction of the energy absorbed by water, dust and metallic structures, of the heat losses and of the average heat transfer coefficient were determined considering pure steam, steam-dust and steam-air-dust injection. The average heat transfer coefficient, determined calculating the steam jet surfaces by means of image elaboration, was compared with empirical correlations.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"11 2","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-11-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139261260","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 Monte Carlo neutron transport method is used to accurately estimate various quantities, such as k-eigenvalue and integral neutron flux. However, in the case of estimating a distribution of a desired quantity, the Monte Carlo method does not typically provide continuous distribution. Recently, the functional expansion tally (FET) and kernel density estimation (KDE) methods have been developed to provide a continuous distribution of a Monte Carlo tally. In this paper, we propose a method to estimate a continuous distribution of a quantity in all phase-space variables using a fully connected feedforward artificial neural network (ANN) model with Monte Carlo-based training data. As a proof of concept, a continuous distribution of iterated fission probability (IFP) was estimated by ANN models in two distinct fissile systems. The ANN models were trained on the training data created using the Monte Carlo IFP method. The estimated IFP distributions by the ANN models were compared with the Monte Carlo-based data that include the training data. Additionally, the IFP distributions by the ANN models were also compared with the adjoint angular neutron flux distributions obtained with the deterministic neutron transport code PARTISN. The comparisons showed varying degrees of agreement or discrepancy; however, it was observed that the ANN models learned the general trend of the IFP distributions from the Monte Carlo-based training data.
{"title":"Estimation of Continuous Distribution of Iterated Fission Probability Using an Artificial Neural Network with Monte Carlo-Based Training Data","authors":"Delgersaikhan Tuya, Yasunobu Nagaya","doi":"10.3390/jne4040043","DOIUrl":"https://doi.org/10.3390/jne4040043","url":null,"abstract":"The Monte Carlo neutron transport method is used to accurately estimate various quantities, such as k-eigenvalue and integral neutron flux. However, in the case of estimating a distribution of a desired quantity, the Monte Carlo method does not typically provide continuous distribution. Recently, the functional expansion tally (FET) and kernel density estimation (KDE) methods have been developed to provide a continuous distribution of a Monte Carlo tally. In this paper, we propose a method to estimate a continuous distribution of a quantity in all phase-space variables using a fully connected feedforward artificial neural network (ANN) model with Monte Carlo-based training data. As a proof of concept, a continuous distribution of iterated fission probability (IFP) was estimated by ANN models in two distinct fissile systems. The ANN models were trained on the training data created using the Monte Carlo IFP method. The estimated IFP distributions by the ANN models were compared with the Monte Carlo-based data that include the training data. Additionally, the IFP distributions by the ANN models were also compared with the adjoint angular neutron flux distributions obtained with the deterministic neutron transport code PARTISN. The comparisons showed varying degrees of agreement or discrepancy; however, it was observed that the ANN models learned the general trend of the IFP distributions from the Monte Carlo-based training data.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"9 4","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135589854","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}
Abstract Fluoride-salt-cooled high temperature reactor (FHR) effectively combines the solid fuel and moderator design of high-temperature gas-cooled reactor (HTGR) technology with the fluoride salt coolant (LiF-BeF2, FLiBe) of molten salt reactor (MSR) technology, enabling low-pressure (∼1 atm, 101.325 kPa), and high-temperature (∼700 °C) operations. The design and operational features of the FHR make it a potentially attractive option for a small modular reactor (SMR), provided that it can be modified and made physically small and operate at a low-enough power level (<350 MWth/<150 MWel). Most FHR-SMR designs use high-assay low enriched uranium (HALEU) fuel in the form of tri-structural isotropic (TRISO) fuel particles, combined with the use of a graphite moderator. However, there are alternative design concepts for an FHR-SMR that may offer superior performance characteristics, while utilizing an alternative fissile fuel supply option. In this exploratory study, lattice physics calculations were performed with Serpent to evaluate an alternative FHR-SMR prismatic fuel block design concept using coated annular fuel pellets instead of TRISO-particle fuel compacts, along with the use of hydrogen-based solid moderator rods made of 7LiH. In initial studies, it was found that fuel blocks with 120 moderator rods made of 7LiH tended to have large positive temperature reactivity coefficients (TRCs), which is undesirable for safety reasons. However, reducing the number of moderator rods to 90 or 54, while increasing the number of fuel rods and coolant holes led to low or negative temperature coefficients. For a prismatic fuel block design with 54-7LiH moderator rods, the isothermal temperature coefficient of reactivity (Isothermal TRC), with simultaneous changes in the fuel (F), graphite (G), hydrogen (H), and coolant (C) temperatures, ranges between −0.159 mk/K and −0.054 mk/K, depending on the operating temperature and fuel burnup. Such alternative FHR-SMR fuels could achieve a single-batch core life of ∼10 years with low enriched uranium (LEU) fuel, and ∼45 years with HALEU, in a 350-MWth reactor core.
{"title":"Assessment of Reactor Physics Characteristics of Prismatic Fuel Concepts with a Hydrogen-based Moderator for Use in a Fluoride Salt-Cooled Small Modular Reactor","authors":"Huiping Yan, Blair Bromley","doi":"10.1115/1.4063388","DOIUrl":"https://doi.org/10.1115/1.4063388","url":null,"abstract":"Abstract Fluoride-salt-cooled high temperature reactor (FHR) effectively combines the solid fuel and moderator design of high-temperature gas-cooled reactor (HTGR) technology with the fluoride salt coolant (LiF-BeF2, FLiBe) of molten salt reactor (MSR) technology, enabling low-pressure (∼1 atm, 101.325 kPa), and high-temperature (∼700 °C) operations. The design and operational features of the FHR make it a potentially attractive option for a small modular reactor (SMR), provided that it can be modified and made physically small and operate at a low-enough power level (&lt;350 MWth/&lt;150 MWel). Most FHR-SMR designs use high-assay low enriched uranium (HALEU) fuel in the form of tri-structural isotropic (TRISO) fuel particles, combined with the use of a graphite moderator. However, there are alternative design concepts for an FHR-SMR that may offer superior performance characteristics, while utilizing an alternative fissile fuel supply option. In this exploratory study, lattice physics calculations were performed with Serpent to evaluate an alternative FHR-SMR prismatic fuel block design concept using coated annular fuel pellets instead of TRISO-particle fuel compacts, along with the use of hydrogen-based solid moderator rods made of 7LiH. In initial studies, it was found that fuel blocks with 120 moderator rods made of 7LiH tended to have large positive temperature reactivity coefficients (TRCs), which is undesirable for safety reasons. However, reducing the number of moderator rods to 90 or 54, while increasing the number of fuel rods and coolant holes led to low or negative temperature coefficients. For a prismatic fuel block design with 54-7LiH moderator rods, the isothermal temperature coefficient of reactivity (Isothermal TRC), with simultaneous changes in the fuel (F), graphite (G), hydrogen (H), and coolant (C) temperatures, ranges between −0.159 mk/K and −0.054 mk/K, depending on the operating temperature and fuel burnup. Such alternative FHR-SMR fuels could achieve a single-batch core life of ∼10 years with low enriched uranium (LEU) fuel, and ∼45 years with HALEU, in a 350-MWth reactor core.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"399 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134957239","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}
Harvey Oliver Plows, Jinfeng Li, Marcus Dahlfors, Marat Margulis
With the ever-increasing role that nuclear power is playing to meet the aim of net zero carbon emissions, there is an intensified demand for understanding the thermal hydraulic phenomena at the heart of current and future reactor concepts. In response to this demand, the development of high-resolution flow analysis instrumentation is of increased importance. One such under-utilised and under-researched instrumentation technology, in the context of fluid flow analysis, is fibre Bragg grating (FBG)-based sensors. This technology allows for the construction of simple, minimally invasive instruments that are resistant to high temperatures, high pressures and corrosion, while being adaptable to measure a wide range of fluid properties, including temperature, pressure, refractive index, chemical concentration, flow rate and void fraction—even in opaque media. Furthermore, concertinaing FBG arrays have been developed capable of reconstructing 3D images of large phase structures, such as bubbles in slug flow, that interact with the array. Currently a significantly under-explored application, FBG-based instrumentation thus shows great potential for utilisation in experimental thermal hydraulics; expanding the available flow characterisation and imaging technologies. Therefore, this paper will present an overview of current FBG-based flow characterisation technologies, alongside a systematic review of how these techniques have been utilised in nuclear thermal hydraulics experiments. Finally, a discussion will be presented regarding how these techniques can be further developed and used in nuclear research.
{"title":"Flow Characterisation Using Fibre Bragg Gratings and Their Potential Use in Nuclear Thermal Hydraulics Experiments","authors":"Harvey Oliver Plows, Jinfeng Li, Marcus Dahlfors, Marat Margulis","doi":"10.3390/jne4040042","DOIUrl":"https://doi.org/10.3390/jne4040042","url":null,"abstract":"With the ever-increasing role that nuclear power is playing to meet the aim of net zero carbon emissions, there is an intensified demand for understanding the thermal hydraulic phenomena at the heart of current and future reactor concepts. In response to this demand, the development of high-resolution flow analysis instrumentation is of increased importance. One such under-utilised and under-researched instrumentation technology, in the context of fluid flow analysis, is fibre Bragg grating (FBG)-based sensors. This technology allows for the construction of simple, minimally invasive instruments that are resistant to high temperatures, high pressures and corrosion, while being adaptable to measure a wide range of fluid properties, including temperature, pressure, refractive index, chemical concentration, flow rate and void fraction—even in opaque media. Furthermore, concertinaing FBG arrays have been developed capable of reconstructing 3D images of large phase structures, such as bubbles in slug flow, that interact with the array. Currently a significantly under-explored application, FBG-based instrumentation thus shows great potential for utilisation in experimental thermal hydraulics; expanding the available flow characterisation and imaging technologies. Therefore, this paper will present an overview of current FBG-based flow characterisation technologies, alongside a systematic review of how these techniques have been utilised in nuclear thermal hydraulics experiments. Finally, a discussion will be presented regarding how these techniques can be further developed and used in nuclear research.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"18 6","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135170862","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}
Abstract I would like to extend New Year's greetings to the readers, reviewers and editors of the ASME Journal of Nuclear Engineering and Radiation Science (NERS) as Chair of the Power and Energy Systems Division (PESD) of the Japan Society of Mechanical Engineers (JSME).
{"title":"Greetings From the Chair of the Jsme Power Energy Systems Division","authors":"Takashi Kiga","doi":"10.1115/1.4063882","DOIUrl":"https://doi.org/10.1115/1.4063882","url":null,"abstract":"Abstract I would like to extend New Year's greetings to the readers, reviewers and editors of the ASME Journal of Nuclear Engineering and Radiation Science (NERS) as Chair of the Power and Energy Systems Division (PESD) of the Japan Society of Mechanical Engineers (JSME).","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"41 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135512849","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}
Abstract A Level 3 Probabilistic Safety Assessment (L3 PSA) is required in UK Generic Design Assessment (GDA) to demonstrate that a new nuclear power plant is suitable to be built in UK. L3 PSA is used to assess the individual and societal risk and compare the results against the offsite Radiation Protection Targets (RPTs) for fault and accident conditions. There is little relevant good practice and mature standard for L3 PSAs that have recently been implemented worldwide. In this study, a pilot L3 PSA is performed for UK HPR1000 to reflect the UK context and relevant good practices. It introduces the methodology and the processes to be followed to perform conditional consequences calculations for the faults and accident scenarios. All radiation sources are considered and analyzed. The radiological risks to a potential UK site are analyzed and compared against RPTs. A widely used code - PC COSYMA, is selected for quantification. The strengths and limitations of the code are justified based on the project situation, and either qualitative arguments or supplementary analysis is subsequently proposed to overcome the limitations. The final L3 PSA results are derived to support the demonstration that the offsite radiological risks for UK HPR1000 have been achieved as low as reasonably practicable (ALARP) and has met the UK regulatory expectation.
{"title":"Application of Level 3 Probabilistic Safety Analysis In UK HPR1000","authors":"Jinkai Wang, Qi Wang","doi":"10.1115/1.4063874","DOIUrl":"https://doi.org/10.1115/1.4063874","url":null,"abstract":"Abstract A Level 3 Probabilistic Safety Assessment (L3 PSA) is required in UK Generic Design Assessment (GDA) to demonstrate that a new nuclear power plant is suitable to be built in UK. L3 PSA is used to assess the individual and societal risk and compare the results against the offsite Radiation Protection Targets (RPTs) for fault and accident conditions. There is little relevant good practice and mature standard for L3 PSAs that have recently been implemented worldwide. In this study, a pilot L3 PSA is performed for UK HPR1000 to reflect the UK context and relevant good practices. It introduces the methodology and the processes to be followed to perform conditional consequences calculations for the faults and accident scenarios. All radiation sources are considered and analyzed. The radiological risks to a potential UK site are analyzed and compared against RPTs. A widely used code - PC COSYMA, is selected for quantification. The strengths and limitations of the code are justified based on the project situation, and either qualitative arguments or supplementary analysis is subsequently proposed to overcome the limitations. The final L3 PSA results are derived to support the demonstration that the offsite radiological risks for UK HPR1000 have been achieved as low as reasonably practicable (ALARP) and has met the UK regulatory expectation.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"24 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":"135513156","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}
Abstract The differential scanning calorimetry (DSC) method has recently emerged as a sophisticated and precise technique for promising contributions to the thermal analysis of various materials, including heavy liquid metal (HLM) coolants. However, there is a lack of experimental studies on the thermal properties of lead-based fluids, such as lead–bismuth eutectic (LBE) and lead–lithium eutectic, which are potential candidates for use as coolants, breeders, and neutron multipliers in advanced nuclear systems like the fourth-generation lead-cooled fast reactor. The available experimental data on the thermal properties of LBE and other lead-based fluids is limited, and the measurements have significant uncertainty. In addition, the composition of components used in the previous studies is inconsistent, and the environmental conditions were often unknown. Therefore, to fill these gaps and advance the thermal properties measurement technique for heavy liquid metal coolants, ENEA Brasimone, in collaboration with DICI-UNIPI, has installed a DSC instrument setup. The experiments performed at the installed DSC setup are focused on measuring some essential thermal properties of LBE using DSC. The experience gained from this work will facilitate the measurement of other fluids based on lead alloy, especially lead–lithium eutectic, a potential candidate for breeder, coolant, and neutron multiplier in demonstration power plant fusion reactors. This study represents the first effort to advance the DSC approach for accurately measuring the thermal characteristics of heavy liquid metals that are highly reactive, such as lead–lithium, which has significant potential in advanced nuclear systems.
{"title":"Advancement Towards the Experimental Measurement of the Lbe Thermal Properties Using DSC Technique","authors":"Satya Saraswat, Nicola Forgione, Massimo Emilio Angiolini","doi":"10.1115/1.4063572","DOIUrl":"https://doi.org/10.1115/1.4063572","url":null,"abstract":"Abstract The differential scanning calorimetry (DSC) method has recently emerged as a sophisticated and precise technique for promising contributions to the thermal analysis of various materials, including heavy liquid metal (HLM) coolants. However, there is a lack of experimental studies on the thermal properties of lead-based fluids, such as lead–bismuth eutectic (LBE) and lead–lithium eutectic, which are potential candidates for use as coolants, breeders, and neutron multipliers in advanced nuclear systems like the fourth-generation lead-cooled fast reactor. The available experimental data on the thermal properties of LBE and other lead-based fluids is limited, and the measurements have significant uncertainty. In addition, the composition of components used in the previous studies is inconsistent, and the environmental conditions were often unknown. Therefore, to fill these gaps and advance the thermal properties measurement technique for heavy liquid metal coolants, ENEA Brasimone, in collaboration with DICI-UNIPI, has installed a DSC instrument setup. The experiments performed at the installed DSC setup are focused on measuring some essential thermal properties of LBE using DSC. The experience gained from this work will facilitate the measurement of other fluids based on lead alloy, especially lead–lithium eutectic, a potential candidate for breeder, coolant, and neutron multiplier in demonstration power plant fusion reactors. This study represents the first effort to advance the DSC approach for accurately measuring the thermal characteristics of heavy liquid metals that are highly reactive, such as lead–lithium, which has significant potential in advanced nuclear systems.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"76 6","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135513941","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}
Abstract During the manufacturing and installation of nuclear power equipment and components, welding-related nonconformance reports (NCRs) prepared by initiators for evaluation vary considerably in quality, influencing working efficiency and project progress. In accordance with the project practices, this article tries to determine the reasons for low-quality welding-related NCR content from three aspects: document flow process, skill levels of technicians, and nonconforming-item control procedure. The technical evaluation requirements on welding-related NCRs are put forward from the perspective of design evaluation in accordance with regulations, standards, design principles, professional knowledge, and engineering practices. According to human performance theory, these evaluation requirements can be used to guide the preparation and review of welding-related NCRs. The inherent logic behind these requirements is to help the initiators and reviewers of welding-related NCRs transform their knowledge-based behavior modes to rule-based behavior modes, thus greatly improving the content quality of NCRs.
{"title":"Technical Evaluation Requirements on Welding-Related Nonconformance Reports","authors":"Chongzhi Wu, Mingkai Li, Zongkui Wang, Zhenguo Peng","doi":"10.1115/1.4063630","DOIUrl":"https://doi.org/10.1115/1.4063630","url":null,"abstract":"Abstract During the manufacturing and installation of nuclear power equipment and components, welding-related nonconformance reports (NCRs) prepared by initiators for evaluation vary considerably in quality, influencing working efficiency and project progress. In accordance with the project practices, this article tries to determine the reasons for low-quality welding-related NCR content from three aspects: document flow process, skill levels of technicians, and nonconforming-item control procedure. The technical evaluation requirements on welding-related NCRs are put forward from the perspective of design evaluation in accordance with regulations, standards, design principles, professional knowledge, and engineering practices. According to human performance theory, these evaluation requirements can be used to guide the preparation and review of welding-related NCRs. The inherent logic behind these requirements is to help the initiators and reviewers of welding-related NCRs transform their knowledge-based behavior modes to rule-based behavior modes, thus greatly improving the content quality of NCRs.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135569164","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}
Abstract This is a ASME Nuclear Engineering Division Chair's editorial message for the January 2024 issue.
这是美国机械工程师协会(ASME)核工程部门主席在2024年1月号上的社论信息。
{"title":"Ned Chair's Message","authors":"Asif Arastu","doi":"10.1115/1.4063844","DOIUrl":"https://doi.org/10.1115/1.4063844","url":null,"abstract":"Abstract This is a ASME Nuclear Engineering Division Chair's editorial message for the January 2024 issue.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"54 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135779307","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}
Abstract The feasibility of transuranic (TRU) fuel rods has been studied for a dual-cooled VVER-1000 assembly along with the conventional UO2 fuel rods. It has been found that the heterogeneous arrangements of TRU and UO2 fuel rods help to improve the multiplication factor, enhance the fuel cycle parameters and maintain the negative Doppler coefficient of the reference VVER-1000 assembly. Within the different heterogeneous combinations of TRU and UO2 fuel rods, the equal number of UO2 and TRU fuel rods, which is referred to as Model 6 increased the multiplication factor of the reference assembly by 7.66% at the beginning of the cycle. Furthermore, the fuel cycle parameter becomes almost double for this Model in comparison with the reference assembly. TRU rods make the neutron spectrum of reference assembly slightly harder and increase the fast fission rate. However, the Doppler coefficient becomes less negative, and flux level decreases with increasing the number of TRU rods. By considering the safety parameters and neutronic behavior, Model 6 (equal number of UO2 and TRU fuel rods) is to be effective for the annular fueled VVER-1000 reactor.
{"title":"Neutronic Analysis and Fuel Cycle Parameters of Transuranic-UO2 Fueled Dual Cooled VVER-1000 Assembly","authors":"Md. Tanvir Ahmed, Afroza Shelley, Md. Nobir Hosen","doi":"10.1115/1.4063757","DOIUrl":"https://doi.org/10.1115/1.4063757","url":null,"abstract":"Abstract The feasibility of transuranic (TRU) fuel rods has been studied for a dual-cooled VVER-1000 assembly along with the conventional UO2 fuel rods. It has been found that the heterogeneous arrangements of TRU and UO2 fuel rods help to improve the multiplication factor, enhance the fuel cycle parameters and maintain the negative Doppler coefficient of the reference VVER-1000 assembly. Within the different heterogeneous combinations of TRU and UO2 fuel rods, the equal number of UO2 and TRU fuel rods, which is referred to as Model 6 increased the multiplication factor of the reference assembly by 7.66% at the beginning of the cycle. Furthermore, the fuel cycle parameter becomes almost double for this Model in comparison with the reference assembly. TRU rods make the neutron spectrum of reference assembly slightly harder and increase the fast fission rate. However, the Doppler coefficient becomes less negative, and flux level decreases with increasing the number of TRU rods. By considering the safety parameters and neutronic behavior, Model 6 (equal number of UO2 and TRU fuel rods) is to be effective for the annular fueled VVER-1000 reactor.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"305 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135803736","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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