bluuebunny/crossref_metadata_2025_split huggingface dataset:
Atomistic and Theoretical Insights on Nanoprecipitates-Controlled Strengthening Behavior for Optimum Mechanical Performance in Nanotwinned NiCoCr Alloys
Authors: Yu Zihao , Zhang Yujie , Wang Hongyu , Sun Ligang , Guo Xiang , Zhu Linli 
Published: November 2024

Abstract
The nanoprecipitates and nanotwins enable to improve the mechanical performance of NiCo-based alloys. In this work, the molecular dynamics (MD) simulations are performed to investigate the strengthening mechanisms of nanotwinned medium-entropy NiCoCr alloys with various distributions and volume fractions of nanoprecipitates. MD simulations reveal that mechanical performance for the precipitates located in twin boundaries is better than that located in the twin lamellae. The precipitate-induced strengthening makes the nanotwinned NiCoCr alloys to achieve the maximum flow stress during increasing the precipitate volume fraction. The influences of volume fraction and distribution of the precipitate on winding and cutting mechanisms are analyzed comprehensively. The dislocation winding behavior, hindered twin boundaries deformation, and the adjacent precipitates connection control the precipitate strengthening mechanisms. A dislocation-based theoretical model is developed to forecast the size-dependent flow stress of nanotwinned metals with nanoprecipitates, in which the Orowan bypass mechanism and the dislocation pile-up behaviors are involved. The relationship between the microstructural size and the flow stress of nanotwinned metallic materials with nanoprecipitates is explored. The predictions for the flow stresses varied with the precipitate volume fraction are agreeable well with the results of MD simulation. The predicted maximum flow stresses and the corresponding critical volume fractions of nanoprecipitates are sensitive to the microstructural sizes.
Design and Development of Scissorbot: A Novel Mid-Flight-Span-Reducing Quadcopter
Authors: Kulkarni Eeshan , Sundaram Suresh 
Published: November 2024

Abstract
This article presents the design and development of Scissorbot, a novel mid-flight reconfigurable geometry quadcopter that reduces its lateral-span using a single servomotor coupled with a compact bevel differential gearbox. Scissorbot possesses unique practical features, including geometrical symmetricity, fault tolerance to the servomotor, and the gearbox’s weight-scalability. Scissorbot achieves significant lateral-span reduction without the risk of propeller tip collision by positioning adjacent propellers in different planes. To the best of author’s knowledge, the maximum lateral-span reduction surpasses any controllable morphing quadcopter reported in the literature. This work derives a detailed attitude dynamics model and conducts a theoretical analysis of the gearbox. The attitude control is accomplished through the implementation of a robust controller, which exhibits exponential tracking, even in the presence of parametric uncertainties, and propeller aerodynamics disturbances. A velocity controller augments the attitude controller. The control allocation loop is parametrized to adapt to the reconfiguration process. Various evaluations, encompassing multibody simulations and hardware flight experiments, assess Scissorbot’s performance. The results unequivocally demonstrate controller’s excellent tracking in simulations and on the hardware prototype across all scenarios. Moreover, the experiments validate the fault tolerance feature of Scissorbot. In addition, a comprehensive study is performed to analyze the scalability of the current gearbox for higher-weight Scissorbot prototypes.
Comparison of Theoretical Methods Describing the Heat Transfer in Vertical Tube Condensers Under Conditions Corresponding to the Condensation of Flue Gas From a Biomass Boiler
Authors: Krempaský Jakub , Havlík Jan , Dlouhý Tomáš 
Published: November 2024

Abstract
A theoretical and experimental study was conducted on the condensation of water vapor in a vertical tube condenser under conditions corresponding to the condensation of flue gas from a biomass boiler, where flue gas is considered a mixture of water vapor and a high content of noncondensable gas (NCG). Four fundamental theoretical methods were identified for determining the heat transfer coefficient and condenser heat output—empirical correlations, heat and mass transfer analogy, diffusion layer theory, and boundary layer theory. These methods were compared in terms of their usability in the design of flue gas condensers in energy systems and experimentally verified. Experiments were carried out in a 1.5 m long vertical double-pipe condenser with a condensing gas flowing in a downward direction through an inner tube with a diameter of 25 mm. The mass concentration of NCG in a mixture with water vapor ranged from 11 vol% to 86 vol% and the inlet Reynolds number of the condensing gas ranged from 1936 to 14,408, which corresponds with the conditions of condensing flue gas from biomass boilers. The boundary layer theory is highly complex and impractical for the calculation of heat exchangers. Empirical correlations have a wide dispersion of the result, because they consider only the fundamental parameters of the process. Nevertheless, heat and mass transfer analogy and diffusion layer theory seem to be the most suitable for flue gas condensers since they capture the physical essence of the phenomena.
Analyzing Flow Rate Impact on CHF Front Behavior During Boiling Crisis in Downward Flow Boiling
Authors: Wang Shixian , Wang Kai , Miwa Shuichiro , Okamoto Koji 
Published: November 2024

Abstract
Our research thoroughly examines the Critical Heat Flux (CHF) in downward-facing flow boiling, essential for comprehending the intricacy of the CHF phenomenon. Conducted on a downward-facing copper surface within a flow loop, our research uncovered significant differences in CHF between upstream and downstream regions, with CHF predominantly occurring downstream first. High-speed camera analysis during boiling crisis led to the identification of the ‘CHF front,’ marking the transition from nucleate to film boiling phases. This CHF front typically originates downstream and progresses upstream, significantly influencing final CHF thresholds. This movement, linked to dry spot formation on the heated surface as per dry-out theories, is a key focus of our analysis. Furthermore, our experiments on smooth copper surfaces under varied flow rates revealed that increased flow rates decrease the CHF front velocity, delaying CHF onset and resulting in higher final CHF values. This observed effect is attributed to the enhanced supply of liquid to the micro-layer beneath the bubble, which increases its thickness and thereby prevents the formation of dry spots. While our study offers deep insights, we recognize uncertainties in our data. Future research will focus on various factors for a more comprehensive understanding of CHF front velocity. These findings are crucial for predicting CHF levels in various systems, marking a significant contribution to the field.
Dynamic Matrix Control for Cold Helium Temperature of Multi-Modular High Temperature Gas-Cooled Reactor Plant
Authors: Zhendong Wu , Zhe Dong 
Published: November 2024

Abstract
The coordination control strategy for operational multi-modular high temperature gas-cooled reactors primarily considers process variables such as neutron flux, main steam temperature, and hot helium temperature, yet it does not address the control of cold helium temperature. Inappropriate cold helium temperatures can lead to structural material fatigue, impact the lifespan of instrumentation at the core’s top, and consequently degrade the operational reliability of nuclear power plants. Additionally, cold helium temperature reflects the thermal balance between the primary and secondary loops, underscoring the necessity of controlling cold helium temperature.
In response, a method using dynamic matrix control (DMC) is proposed to control cold helium temperature by adjusting the hot helium temperature setpoint. A step response model with cold helium temperature data is used to design DMC. Utilizing the stability of the existing proportional-integral-derivative (PID) controller and the optimization capabilities of DMC, a cascaded control structure is established. PID acts as the inner loop for stability, and DMC as the outer loop corrects the hot helium temperature setpoint, achieving cascaded control for cold helium temperature in multi-modular high temperature gas-cooled reactors.
Numerical simulations based on a 1000MW ten-module high temperature gas-cooled reactor program validated the effectiveness of the proposed control method. The results demonstrate that the controller can effectively regulate the cold helium temperature without compromising the control of other key variables.
Simulation Calculation of the Utilization Rate of Uranium Resources on the ADANES Fuel Cycle
Authors: Yan Xuesong , Gao Yucui , Zhang Yaling , Yang Lei 
Published: November 2024

Abstract
Accelerator Driven Advanced Nuclear Energy System (ADANES) is composed of Accelerator Driven subcritical System (ADS) burner and regeneration system of spent fuel, which can solve the challenging problems of low utilization efficiency of uranium resources and difficult safe disposal of spent fuel. The proportion of U-238 in spent fuel of pressurized water reactors is ∼95%, fission products are 3∼4%, and other actinide nuclides are 1∼2%. The regeneration system of spent fuel is mainly to eliminate part of fission products, including volatile, semi-volatile and rare earth nuclides, remaining nuclides to form regenerated spent fuel (mainly including actinides and other fission products). It does not perform fine separation. ADS burner can burn the regenerated spent fuel, so that ADANES can realize power generation, transmutation and proliferation for multiple fuel cycles. ADANES aims to increase the utilization rate of uranium resources to 95% and generate very little radioactive waste with a short radioactive life. At present, the preliminary principle verification device of the burner Chinese Initial ADS has been successfully applied and will be constructed in 2027. In this work, the neutron physics effects of these nuclides (volatile, semi-volatile and rare earth nuclides) on the ADS reactor core will be studied. The ADS burner can directly utilize regenerated spent fuel because of its powerful external neutron source. After removing 100% of volatile, semi-volatile and more than 50% of rare earth nuclides from the spent fuel of PWR, 3GWth ADS burner can burn the remaining spent fuel for 10∼20 years. The Keff of the reactor core will first increase to its peak and then continuously decrease, this is due to the production of plutonium-239, the consumption of uranium-235 and production and accumulation of neutron toxins. Under subcritical operating conditions of the reactor, the mass of U-235 will consume &gt; 99% and U-238 will consume &gt; 70% during 100 years. Meanwhile, a large amount of Pu-239 has also proliferated. The generated Pu-239 is equivalent to the consumed U-235. We attempt to calculate the number of cycles and fuel utilization on a feasible basis. Specific analysis was conducted on the requirements for initial Keff of ADS reactor core and accelerator beam intensity. More the 5 regeneration system and nuclear fuel cycle will be simulated. As the number of cycles increases, the utilization of U-238 will continue to increase. After all, there will be more and more fission products accumulated in spent fuel, which will affect the burnup level.
A Method Using Strong Tracking Filter for Compensating the Response Delay of Self-Powered Neutron Detectors
Authors: Liu Shiyu , Zhang Qingmin , Liu Jinliang , Sang Yaodong , Shao Zhuang , Zhu Kangfu 
Published: November 2024

Abstract
Self-Powered Neutron Detector (SPND) has been widely used for monitoring neutron flux in reactors. There are two typical types of SPNDs, including prompt type and delayed type. The prompt type SPNDs’ output current are mainly contributed by gammas generated from the neutron capture process.
Rhodium SPND is a typical delayed type SPND and has been widely used for neutron flux monitoring in reactors. Compared with other prompt SPNDs such as Cobalt SPND, Rh-SPND contains a higher neutron sensitivity.
However, long response time and high noise gain are the main problems that restrict the application of SPNDs with delayed current components. In order to solve the above two problems, a method using the Strong Tracking Filter has been proposed in this paper. Compared to the traditional Kalman Filter, Strong Tracking Filter will offer robustness to the modeling errors and ensure the whiteness property of the filter innovation sequence. This gives Strong Tracking Filtering a high adaptivity to sudden changes (e.g. reactor load rejection) in the useful signal, even in the presence of noise. Following the derivation of the ideal iterative inversion relation from current to neutron flux, a filtering matrix containing noise filtering and neutron flux inversion was established using Strong Tracking Filter. Then it’s shown that this method can give a shorter response time than the Kalman Filter method for the same noise gain, and its strong robustness to initial values and sampling frequency was also demonstrated by the test. Finally, this method was used to obtain the real-time neutron flux from the SPND current in the other reference, demonstrating its effectiveness and excellence.
Transient Analysis of the Loss of Heat Sink Accident in a New Type of Megawatt Heat Pipe Reactor
Authors: ZhenLan Wang , JunLi Gou , LeQi Yuan , JianQiang Shan 
Published: November 2024

Abstract
Heat pipe reactors are one of the ideal reactor types for Unmanned Underwater Vehicles (UUVs) due to high energy density, long lifecycle, modularity, and compact structure. In the reactor concept design stage, simulation and analysis of typical accidents on heat pipe reactors are key to assessing their inherent safety. The loss of heat sink accident, as one of the typical accidents in heat pipe reactors, is of great significance to the safety performance evaluation. In this study, the transient models of the passive residual heat removal system (PRHRS), including the one-dimensional flow model, the constitutive model of heat transfer, and the constitutive model of pressure drop, are developed based on the completed design of the PRHRS. Based on the models, a transient coupling analysis code for the loss of heat sink accidents is developed, which combines with the transient analysis code for heat transfer in a new type of megawatt heat pipe reactor (HTR) and the transient analysis code for the natural circulation of cooling water (SCTRAN). The results of the coupling analysis code are validated against the results of the STAR-CCM+ during the steady-state operation, with the maximum error not exceeding 8.13%. The coupling analysis code is also used to analyze the transient characteristic of the PRHRS during the loss of heat sink accident with different delayed emergency reactor shutdown times. After the loss of heat sink accident, the cooling water in the emergency cooling component (ECC) is natural flow driven by density differences without external power. Before the reactor is shut down, the heat generated in the core can only be removed through the PRHRS, and the temperature of the core remains to rise. As the heat removal capacity of the PRHRS increases and the reactor shuts down, the fuel and heat pipes gradually cool down after reaching the peak temperature, and the waste heat of the core is gradually removed. The results show that with the delayed emergency reactor shutdown time within 23s after the loss of heat sink accident, the PRHRS can remove the residual heat of the core in a timely manner, no boiling occurs in the ECC, and the temperatures of the core are lower than the temperature limit, which reflects the good safety of the new type of megawatt heat pipe reactor in the accident of the loss of heat sink.
Capability for Volume Reduction of Concrete Contaminated by Radioactive Carbon Dioxide Using Rubbing
Authors: Kinoshita Norikazu , Nakashima Hitoshi , Saito Akira , Hanzawa Mamoru , Sasaki Yuki , Torii Kazuyuki 
Published: November 2024

Abstract
A large amount of concrete contaminated by 14CO2 will be discharged from decommissioning of aged nuclear power plants. Rubbing, which separates cements from concrete debris, would be one of manners which allows to reduce volume of the waste if most of the 14C presents in the cement inside the concrete. We confirmed that the 14CO2 adsorbs on the cements more than aggregates by a factor of approximately 2 orders. Rubbing was carried out to obtain the mass balance and decontamination factor using simulated concrete debris which was not contaminated by the 14CO2. The cements and fine aggregates were separated from the debris as fines by rubbing using a Los Angeles testing machine. Steel balls with different size and different amount were used to proceed rubbing more. In addition, production rates of the fines were investigated in the debris that the heat treatment was carried out or not. Residues in a mill of the Los Angeles testing machine were washed to remove deposits remained on the surface. We concluded that the volume reduction can be effectively achieved by rubbing the debris that applied the heat treatment using more steel balls with larger size. DF was not improved by washing the surface residues. However, washing can resolve the concerns regarding radiation protection due to scattering of fines contaminated with the 14CO2 in treatment of the residues after rubbing.
Study on the Improvement of RELAP5 Post-Critical Boiling Heat Transfer Model for the Submerged Cooling Process of Rectangular Channel
Authors: Zhong Lei , Yu Hongxing , Ling Chen , Chen Deqi , Deng Jian , Liu Haidong , Xu Jun 
Published: November 2024

Abstract
RELAP5 has some shortcomings in calculating the submerged cooling process of rectangular channels, especially in the post critical heat transfer model. Based on the experimental data of submerged cooling in rectangular channel, the droplet flow precursor cooling heat transfer model, transition boiling model and minimum film boiling temperature model suitable for the submerged cooling process in rectangular channel are obtained in this paper. The secondary development is carried out on the basis of RELAP5 program, and the droplet flow precursor cooling heat transfer model, transition boiling model and minimum film boiling temperature model obtained in this paper are implanted. Through the comparative analysis of program calculation results and experimental data, the results show that the calculated values of this model are in good agreement with the experimental data, and compared with the existing calculation models, this model has good applicability to calculate the submerged cooling process of rectangular channel.
Climate Change: Overview of the Process and Methodologies Used by EDF for Taking Into Account the Effects of Climate Change in the Design of Nuclear Power Plants
Authors: Cordier Hervé , Joly Amélie 
Published: November 2024

Abstract
Climate Change is one of the greatest environmental challenges facing the world. Its effects are systemic and evolutive because of its dependance with human activities. The Intergovernmental Panel on Climate Change (IPCC) highlights that the world efforts to reduce greenhouse gases emissions are not on track to limit global warming to 1.5°C by 2100 in comparison to pre-industrial era.
As several countries, France signed a legal duty to reach Net Zero by 2050. Generation of low carbon electricity is part of solutions required to meet the Net Zero target. EDF operating nuclear fleet in France updated its life cycle analysis leading to a carbon footprint of 4 grams of CO2 equivalent per kWh [1]. On one hand, this result highlights that alongside renewables, nuclear facilities are part of the solutions to reduce greenhouse gases emissions. On the other hand, those facilities require to be adapted and resilient to the effects of Climate Change for their entire lifetime.
For Nuclear Facilities, Climate Change can impact safety, electricity generation and environmental impacts during the different steps of the facility development (design, building) and for its entire lifetime (operation and decommissioning).
Considering the possible effects of climate change, EDF has developed a systemic and evolutive approach to design and manage the adaptation process of its nuclear facilities with climate change. This approach called “ADAPT program” is structured around 4 pillars (understand, characterize, mobilize and act) and 4 work packages: (i) natural and water resources, (ii) industrial facility, (iii) supply chain local logistics work conditions and (iv) local environment of the Nuclear Power Plant (NPP).
In this article, we focus on the industrial facility work package. The whole process is described, and the goals of the different steps are explained as well as the principles applied to integrate the effects of climate change in the process. The risk assessment step is detailed. An example for the design against heatwaves is given for new and existing Nuclear Power Plant.
Preliminary Study on Endoscopic Ultrasound System for In-Service Inspection of Helical Tube Once-Through Steam Generator
Authors: Xi Ze , Wang Xiangang , Luo Xiaowei , Zhou Junjie , Song Lei 
Published: November 2024

Abstract
The Helical-tube Once-through Steam Generator (H-OTSG) is a crucial component of the gas-cooled reactor, with the helical heat exchanger tube in the H-OTSG forming an integral part of the reactor’s primary pressure boundary. The volumetric in-service inspection of this helical heat exchanger tube presents a significant challenge due to constraints of accessibility, material, and weld inspection. To address this problem, an Endoscopic Ultrasound (EUS) inspection system is proposed. The structure and working mode of the EUS inspection system are described in detail. Then, the key sensor of the inspection system, EUS transducer, is presented and highlighted. Three array EUS transducers are designed. After outlining the geometric model and inspection modality of the EUS transducers, a preliminary EUS inspection experiment is conducted to demonstrate the inspection performance of the transducers. Two types of reflectors are fabricated on an actual H-OTSG’s heat exchanger tube in the High-Temperature Reactor Pebble-bed Module (HTR-PM) according to ASME code methods to simulate the volumetric and planar defects in different positions. The experimental results show that the proposed array EUS transducers offer high detection robustness and excellent sizing ability on account of the 3-D focused acoustic beam.