Journal of Nuclear Engineering and Radiation Science最新文献

{"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":null,"pages":null},"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}

{"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":null,"pages":null},"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}

{"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":null,"pages":null},"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}

{"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":null,"pages":null},"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}

{"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":null,"pages":null},"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}

{"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":null,"pages":null},"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}

{"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":null,"pages":null},"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}

{"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":null,"pages":null},"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}

{"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 (<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.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"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}

{"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":null,"pages":null},"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}