bluuebunny/crossref_metadata_2025_split huggingface dataset:
General Discussion on Terminal Velocity for Rising Single Bubble
Authors: Wang Qinghua , Kawara Zensaku , Yokomine Takehiko , Kunugi Tomoaki 
Published: October 2021

Abstract
Distinct from solid particle, bubble is soft matter, which tends to adjust its shape according to motion, and according to the work of Tomiyama, for bubble with diameter larger than 1.3mm, the rising path would change from rectilinear to zig-zag or helical, which make the determination of the terminal velocity more difficult. Even for the bubble with diameter larger than 0.075 mm reported by Pawliszak, terminal velocity scatter started.
The existence of surfactants or contaminants and different initial deformations of bubbles introduced by varying capillary or nozzle diameters were generally considered to be the two main reason account for the terminal velocity scatter.
In this paper, the rising characteristics of small bubbles with 1 &lt; Re &lt; 100 were investigated. Especially the motion characteristic right after the departure. Comparisons were made between three bubbles with almost the same diameter (0.58 ± 0.008mm), the same aspect ratio (0.98 ± 0.01), it was found that the only differences lied on oscillation frequencies of aspect ratio. And the higher oscillation frequency of aspect ratio was, the higher steady-state velocity the bubble would achieve. Therefore, it was thought that the deformation is one of the ways for bubble to accelerate itself, i.e., bubbles accelerated itself through the oscillations of aspect ratio. The oscillation of aspect ratio is of high possibility to be one reason for velocity scatter.
Anomaly Detection of Electric Gate Valve Based on Multi-Kernel Support Vector Machine
Authors: Luo Jing , Wang Hang , Peng Minjun 
Published: October 2021

Abstract
Valve is an indispensable fluid control component in nuclear power system. Nuclear power station has a large number of gate valve equipment, which works under high temperature, high pressure, high radioactivity and other harsh conditions. In nuclear power plant accidents and economic losses, a considerable part of them are caused by valve failure. Aiming at the fault of electric gate valve, this paper proposes an anomaly detection method based on multi-kernel support vector machine. Firstly, the acoustic emission instrument is used to measure the fault state data and extract the fault features. Secondly, on the basis of classical support vector machine, multiple kernel function combinations are selected to decompose the model into convex optimization problems to realize the abnormal state detection of internal leakage fault of electric gate valve in nuclear power plant. The results show that, compared with the classical support vector machine method, the constructed support vector machine method based on multikernel learning has better effect and higher accuracy in anomaly detection of electric gate valve.
Comparative Study on Thermal Stress Analysis and Fatigue Curve in Stress and Fatigue Calculation of Nuclear Equipment
Authors: Shao Xuejiao , Xie Hai , Zhang Yixiong , Zhang Liping , Fu Xiaolong , Mi Xue , Li Hui 
Published: October 2021

Abstract
In the stress and fatigue evaluation of primary nuclear equipment, thermal stress calculation and fatigue curve have a great impact on the calculation results. There is a problem of stability in the step integral when calculating the thermal stress. The transient heat conduction changes rapidly near the initial temperature time point. It is critical to choose the right sub steps. In the nuclear code, the design fatigue curve in the air environment is given by applying a factor of 2 on stress or a factor of 20 on cycles. The factors are used to consider various influences such as data scatter, structure size, surface roughness and load. The factors of fatigue curves in ASME BPVC 2007 edition + 2009 addendum and French AFCEN’s RCC-M 2017 edition RPP 2 have changed. Due to the changes of the original test data and modificative factors, the new fatigue curve and the original fatigue curve are crossed. In code case N-761 of ASME, fatigue curves considering the influence of primary coolant environment are given under different strain rates.
In the calculation of thermal transient using commercial ANSYS software, the setting of time step has a great influence on the thermal stress. This paper compares and analyzes the influence of different calculation methods on the thermal stress at different positions of the model. The effect of different fatigue curves on fatigue life is also discussed considering different air design curve and different strain rate fatigue curve in water.
The results show that the fatigue results obtained by using the modified ASME or RCC-M air fatigue curve are smaller than those before the change when the usage factor is large to a certain extent. After considering the environmental fatigue curve, the fatigue results increase significantly. The accuracy of thermal stress calculation can be significantly improved by adding time points and refining time steps near the time points after the transient sharply changes, otherwise the influence of thermal stress on stress and fatigue results will be underestimated obviously.
Extending a Fluid-to-Fluid Similarity Rationale for Heat Transfer at Supercritical Pressure to R134a
Authors: Kassem Sara , Pucciarelli Andrea , Ambrosini Walter 
Published: October 2021

Abstract
In the frame of the development of Gen-IV supercritical water reactors (SCWRs), making use of the experimental data collected so far by different fluids in view of simulating the behavior of water is an important and very useful step. However, conducting experiments with simulant fluids poses the problem of establishing a clear rationale for inferring the expected water behavior from the information gathered with other fluids.
In literature, several attempts have been made to suggest dimensionless groups to be preserved in establishing a similarity. However, difficulties were encountered in getting appreciable success from these attempts and relevant reviews declared a generally unsatisfactory behavior of available theories.
Recently, some of the authors of this paper were able to demonstrate the good success of their proposal for a fluid-to-fluid similarity theory addressing mainly four relevant fluids being water, carbon dioxide, ammonia and R23. In the frame of the EU ECC-Smart Project, the refrigerant R134a will be used as a simulant fluid for performing experimental analyses in support to the achievement of the project objectives. Hence, the present paper is aimed to show the capabilities of the similarity theory when applied to R134a, proposing a rationale to be used in selecting the boundary conditions of experiments in order to make them really representative of the addressed water plant conditions.
Evaluations of TiO2 Deposition on Structure Surfaces and Water Radiolysis for the Corrosive Environment in a Reactor Pressure Vessel
Authors: Mawatari Takashi , Yamamoto Yasushi , Shibasaki Osamu , Hara Takahiro , Horayama Yusuke , Takagi Junichi 
Published: October 2021

Abstract
It is important to avoid a risk of stress corrosion cracking (SCC) of structure materials in nuclear power plants. One of the methods for preventing SCC is titanium-dioxide (TiO2) injection. TiO2 is a photo-catalyst, and injected TiO2 deposits on the surfaces of structure materials, and the deposited TiO2 can decrease the electrochemical corrosion potential (ECP) by irradiation of Cherenkov-light in the reactor. Generally, the ECP is an important indicator about susceptibility to SCC, and ECP-lowering makes a risk of SCC lower. Therefore, initiation and propagation of SCC will be mitigated by the TiO2 injection.
The evaluation model for corrosive environment had been already developed. That “conventional” model was consisted of some analysis such as water radiolysis, ECP and others. However, the model did not consider an effect of TiO2 deposition and was one-dimensional simple model. Therefore, to evaluate ECP distribution in a reactor after TiO2 injection, the “newly developed” model was build: analyses integration, highresolution and three-dimensional model.
In this paper, the TiO2 deposition and water radiolysis analyses integrated into flow analysis were contained. Both the deposition and the radiolysis depend on flow in the RPV, so that these analyses were built into the computational fluid dynamics (CFD) analysis model for CFD code. The target plant type for the model was 1100 MWe-class BWR-5.
The flow analysis was carried out at steady-state, assuming the TiO2 injection during the rated operation of the plant. Then the TiO2 deposition and the radiolysis analyses were conducted by transient calculations using the result of flow, and the three-dimensional and high resolution results were obtained.
Features of a BWR Neutron Absorber Melt Relocation in an Oxidative Environment During the CLADS-MADE-02 Test
Authors: Pshenichnikov Anton , Nagae Yuji , Kurata Masaki 
Published: October 2021

Abstract
The work on the understanding of the accident progression at the units of the Fukushima Dai-Ichi Nuclear Power Plant (1F) is ongoing. This contribution gives a part of detailed investigations of the control blade melt propagation downwards through the prototypic BWR bundle assembly during the CLADS-MADE-02 test, where the conditions of the 1F Unit 3 was simulated. Interesting features emerged in an oxidative environment.
Influence of Fuel Pellets’ Thermal Expansion on Temperature Feedback Regulation for Megawatt-Class Space Gas-Cooled Fast Reactor
Authors: He Yuhao , Yuan Dongdong , Ning Kewei , Wang Xiaoyu , Zhao Fulong , Tan Sichao 
Published: October 2021

Abstract
As human beings explore the outer space, the demand for the nuclear sources in space missions is increasing. Almost all the existing space nuclear reactors have adopted the design of gas-cooled fast reactor, and the temperature feedback effect of the fuel of the small fast neutron reactor is different from that of the traditional pressurized water reactor. For the space fast reactors, they are designed to be compact and with a long operating lifetime to accomplish the long-term space missions. Due to the high-level enrichment of U-235 in the fuel, the Doppler effect of U-238 has little impact on the reactor reactivity. In the present paper, based on the design of the megawatt-class space gas-cooled fast reactor, the overall temperature feedback adjustment capability of the core, and the overall temperature coefficient of the core including fuel, structural materials, and coolant are investigated. The calculation results show that the temperature of the coolant has little effect on the reactivity. The temperature coefficient of the fuel is positive, and the structural material with a large negative temperature feedback coefficient. At the same time, the deformation of the fuel pellets due to thermal stress at high temperatures and the reactivity changes caused by the deformation are analyzed.
Modeling of Irradiation-Induced Thermo-Mechanical Coupling Behavior in Metal Matrix Microencapsulated Fuel
Authors: Wei Hongyang , Zhu Fawen , Ru Jun , Wang Haoyu , Zhang Jing , Li Hua , Li Yun , Huang Chunlan , Li Yuanming , Ding Shurong 
Published: October 2021

Abstract
The application of TRISO-based dispersion fuel to advanced light water reactors, small modular reactors and micro-reactors has attracted increasing attention globally. With combination of superior fission products retention capability of tri-structural isotropic (TRISO) particle and stability of zircaloy matrix, metal matrix microencapsulated (M3) fuel becomes a promising candidate. To guarantee the safety and obtain the maximal fissile loading, it is necessary to predict inpile thermo-mechanical coupling behavior of M3 fuel. In this study, a three-dimensional finite element analysis for thermo-mechanical coupling behavior of M3 fuel at light water reactors (LWRs) operation condition is performed based on ABAQUS platform. Irradiation effects are considered comprehensively, especially kernel-buffer-IPyC interaction is depicted based on the newly developed model. The actual irradiation boundary conditions in light water reactors are adopted based on existing studies. Distribution and evolution of temperature, displacement and stress-strain are obtained and analyzed. Numerical simulation shows that (1) irradiation-induced swelling of fuel kernel can be well accommodated by porous buffer layer and the particle volume increases only 0.039% at 19% FIMA; (2) the maximal first principal stress for the coating layers will not exceed the corresponding fracture strengths, so that the safety of TRISO particle and zircaloy matrix can be guaranteed. This study serves as a basis for advanced fabrication of M3 fuel in the future.
Automated Eddy Current Array Sensor Delivery Tool for Nondestructive Examination of Spent Fuel Pool Liner
Authors: Smith Michael , Abbate Emily , Phillips Joey , Yoon Byungsik 
Published: October 2021

Abstract
Inspection of a spent fuel pool (SFP) liner (e.g., detection of defects in the pool liner) is important to validate the integrity of this component to perform its intended function. Eddy current array (ECA) testing offers a reliable and sensitive nondestructive examination (NDE) technique to inspect SFP liners, but requires a delivery system to transport the ECA sensor to these hard-to-reach and harsh environment locations. This work presents an ECA sensor delivery tool for use with inspection of an SFP. The design for a payload delivery tool proof-of-concept model is proposed that can transport an ECA sensor with positional feedback. The tool allows the sensor to traverse the wall of a nuclear SFP liner, with both horizontal and vertical motion capabilities to support positioning and inspection. The tool provides both manual and automatic operating modes with variable speeds. Additionally, the delivery tool can be adapted to accommodate other NDE sensors (e.g., ultrasound testing probes). Computational based simulations and analysis validate that the performance requirements for the ECA sensor delivery tool were satisfied to support inspection of an SFP liner upper region.
Neutron Physics Characterization & Optimization Analysis of the ACPR100 Small Modular Reactor
Authors: Liu Songyang , Wang Xiang 
Published: October 2021

Abstract
The ACPR100 is a small modular pressurized water reactor design proposed by China General Nuclear Power Corporation which integrates most important components of the reactor into one pressure vessel. This paper aims to model the reactor core by Monte Carlo code Serpent.
Firstly, the steady-state characterization and optimization analysis within different temperature, pin-pitch and other design parameters are performed. Secondly, the loading pattern of fuel assemblies with Gd-doped fuel rods was assessed. Thirdly, based on the reference research, the temperature coefficients of fuel and coolant are calculated. The influence of control rod insertion depth was simulated additionally. Besides, to show the life cycle and the change of inventory directly, we performed burnup calculation based on pre-defined time step and discussed the fission products including radioactive minor actinides and radiopharmaceutical isotopes during the life cycle.
The results show that the ACPR100 achieved more stable performance using high abundance boric acid, and the negative temperature reactivity coefficient is sufficient to maintain the stability of the reactor operation, but the ACPR100 is not suitable for massive production of radiopharmaceutical isotopes.
A Sodium-Cooled Thermal-Spectrum Fission Battery
Authors: McDaniel Patrick J. , Forsberg Charles 
Published: October 2021

Abstract
We propose a low-enriched-uranium, thermal-neutron-spectrum sodium-cooled reactor with a peak sodium temperature of 700°C coupled to an air-Brayton power cycle for electricity and heat. Three low-enriched-uranium, thermal spectrum sodium-cooled reactors were built in the 1960s and 1970s; but there has been no examination of such systems for many decades. We develop a pre-conceptual design based on “new” enabling technologies since the 1970s including yttrium hydride as the high-temperature neutron moderator, commercial gas turbines and secure decay heat removal systems. We define the reactor as a sodium hydride reactor (SHR).
The initial application is as a fission battery. The concept of the fission battery (FB) is a “plug and play” nuclear reactor defined by multiple characteristics: economics enabled by factory fabrication of large numbers of identical units, easy installation and removal, unattended operation and highly reliable operations. FBs are designed to be a low-carbon replacement for fossil fuels by industrial and commercial companies that require energy to produce some product (manufactured goods, chemicals, education, data centers, ship transportation, etc.). The reactors may be owned or leased by the company. The SHR is at an early stage of development.
A Study on Radiation Imaging Mechanism and Characteristics in Different Inspection Systems
Authors: Xu Yuting , Wu Zhifang , Wang Qiang 
Published: October 2021

Abstract
Radiation imaging, as a key issue in nuclear technology, has received considerable attention in the industry. It is widely used in nuclear medicine, Customs supervision, and many other areas. The objective of this investigation is to get insight into the principle, operation characteristics and image characteristics of radiation imaging.
In this paper, an investigation on radiation imaging is conducted on three main inspection systems for Customs supervision, including small X-ray inspection machine, CT baggage inspection system, and large container inspection system. The principle, operation characteristics, evaluation indexes, pseudo-color processing and image characteristics are discussed in detail. The results indicate that the spatial resolution of small X-ray inspection machine is much higher than that of CT baggage/goods inspection system and large container/vehicle inspection system. It is a challenge to identify substances and specific shapes in the case of overlapping for small X-ray inspection system.
Moreover, the mechanism of X-ray images is discussed as well. The radiation images are divided into three types, including two-dimensional, pseudo-color, high spatial resolution; two-dimensional, gray, high spatial resolution; three-dimensional, pseudo-color, high density resolution. The further investigation on machine inspection images is suggested to focus on the application environment. For some objects with specific characteristics, such as amorphous, explosive, the CT baggage inspection has much better performance than other systems.
The research in this paper reveals the mechanism, parametric effect and imaging characteristics. It could provide a necessary foundation for the follow-up intelligent processing, detection, identification and annotation for radiation imaging in nuclear area. The research on inspection devices could lend strong experience to medical treatment, industry and many other fields.
Assessment Of The Interfacial Drag Models In RELAP5 With Mixture Level Swell Experiment
Authors: Zhu Wen , He Qingche , Zhang Luteng , Ma Zaiyong , Sun Wan , Xu Wangtao , Pan Liangming , Huang Tao , Ding Shuhua 
Published: October 2021

Abstract
The two-phase flow phenomenon can be encountered in whether nuclear reactor normal running condition or emerged accidents, and imposes significant influences on reactor safety. One of the essential characteristics of a two-phase flow is the interfacial drag, which can influence the two-phase interface’s deformation, the distribution of the two phases, the flow pressure drop, etc. In this paper, the interfacial drag models in RELAP5 are assessed using GE mixture level swell experiments and FRIGG tests. GE Small Level Swell experiment and GE Large Level Swell experiment were respectively based on small and large experimental sections to study critical flow, void fraction distribution, two-phase level, and other characteristics under reactor blowdown conditions. FRIGG tests, using steam and water as the two-phase working fluid, studied the steady-state distribution characteristics of axial void fraction in the rod bundle channel under different conditions. They are separate effects tests and used to assess the interfacial drag models for many system analysis codes. For the GE Small Level Swell experiment, the results indicate that the RELAP5’s prediction of the void fraction at the middle and top part of the test section are in good agreement. However, at the bottom of the test section, the prediction is much higher than the experimental data and beyond the experimental data’s error bar. For FRIGG tests, the most simulation results of RELAP5 are similar to the simulation results of TRACE. For high flow rate, high-level power, and large inlet subcooling operation conditions, the prediction of the low void fraction region (below 0.3) and the high void fraction region (beyond 0.6) will have some differences comparing with the experimental data. All the differences between simulation results and experimental results can be attributed to the adopted interfacial drag models’ applicability in RELAP5.
Experimental Simulation of Transitions Between Forced Circulation and Natural Circulation With Nuclear Reactivity Feedback
Authors: Chen Hanying , Xia Linzhong , Gao Puzhen , Tan Sichao , Yuan Hongsheng 
Published: October 2021

Abstract
For a nuclear power reactor operating under forced circulation (FC) and natural circulation (NC) conditions, during transitions between these two types of circulation condition, nuclear reactivity fluctuation will occur due to sudden coolant flowrate change, and the reactivity fluctuation will in turn cause variation of nuclear heating power. In order to provide a more achievable and less costly approach for experimentally investigating transient thermal-hydraulic behavior of these transitions between NC and FC in nuclear power reactors, we designed and established an electrically-heated experimental loop. The established experimental loop has a heating section, whose heating power is controlled in real time and programmed to simulate the nuclear power variation caused by fuel and coolant temperature reactivity feedback. This experimental simulation is specifically implemented by coupling the point kinetics model with temperature measurements of the electrically heated experimental loop. In our study, reactivity feedback during both NC-to-FC transition and the reverse transition were simulated experimentally, and different reactivity compensation strategies were tested. The simulation results show that: in transitions from FC to NC, the decrease of power in initial stage of the transitions can restrain the fast increase of the fuel temperature; while in transitions from NC to FC, enough negative reactivity should be inserted as soon as the circulation pump is shut down, so as to avoid the nuclear power increase peak.
Steam Generator Grade P91 Steel Components Creep-Assessment By Test After Extended Service
Authors: Grisolia Ottaviano , Scano Lorenzo , Piccini Francesco , Lo Conte Antonietta , De Agostinis Massimiliano , Fini Stefano 
Published: October 2021

Abstract
Previous study carried out creep analysis for steam generator high-temperature-section two components, outflow tubing and manifold of the superheater harp: they may have been critical because of the long continued service (109,000 hours or twelve years) and loading conditions, including maximum operation temperature (565°C) and applied stress (65 MPa). Metallographic methods by replica had showed no evidence of the creep cavitation in all the positions considered for both tubing and manifold. In particular, they had not found any cavitation or phases affecting creep strength of the material in the base, HAZ and weld metal microstructure. Now, present study carries out investigation for the two components based on the next plant outage outcome, after further 20,000-hours service. Both metallographic methods and hardness measurements’ results would compare with previous ones providing microstructure evolution in the period.