bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Influence of Impurity Water for CHF on the Downward Facing Boiling Surface
Authors: Lin Bo , Yun Xiangyu , Xu Junying , Wei Dongshan , Zhang Huiyong 
Published: November 2022

Abstract
In-Vessel Retention (IVR) strategy is one of the most important severe accident management strategies. The Critical Heat Flux (CHF) is an important key element in the IVR research. How to increase the CHF get more and more attention in the CHF research. This paper aims at the influence of the coolant with different impurity. Those impurity can fall into the coolant when the severe accident happens. In this experiment, the deionized water was used as the coolant. The experiment investigated the different impurity with boronic acid that has different concentration and the boronic acid water with different impurity, such as concrete, fiberglass and coating. The concrete can be fallen from the construction, the fiberglass can be fallen from the insulation cotton which cover the pipe and device, and the coating can be fallen from the devices and other constructions. When the accident happens, those impurity can fall from the construction, then fall into the coolant and influence the CHF of the outside surface of lower head. From the result of the experiment, the CHF with the boron acid water is higher than that only deionized water. The CHF with the impurity water depend on the concentration of the impurity. Lower concentration will increase the CHF when the higher concentration will decrease the CHF. Different impurity has different influence on CHF, we control the component of the coolant to improve the CHF.
Research on Immersive Simulation of Key Equipment Process Flow in Nuclear Power Plant Based on VR Technology
Authors: Zhang Yi , Zhou Zhi-Gang , Wang Hao , Yang Zheng-Hui 
Published: November 2022

Abstract
Nowadays, 3D modeling and virtual reality technology has become a new method to enhance the digitization level in nuclear power engineering field. Reactor pressure vessel head handling, refueling and main equipment installation are the important process flows in nuclear power plant. Traditional measures without digitization and visualization, which need much more experience and technical skills to conduct the staffs on-site construction work. In that case, there are more risks when they do some complicated operations especially the new ones they never know before. This paper presents a technical method to design an immersive virtual environment, which contains the whole 3D model data of the nuclear island buildings in nuclear power plant to simulate the key processing flows. Meanwhile, the paper develops an interactive VR software system to help training workers about their on-site works in a virtual way before they solve on-site problems. On the other hand, another application of this paper’s work is design verification of the process flows themselves, so that the simulation results can tell designer what and where is the optimizing goal. The achievements of this paper have already been applied in the actual HPR1000 nuclear power project of CGN, and bring considerable efficiency return in construction field.
A Comparative Approach of a Seismic Response Analysis Based on Housner Model and Added Mass Model for a Nuclear Water Storage Tank
Authors: Yang Chengzhuo , Metais Thomas , Zheng Zihao , Dai Yang , Min Jiesheng 
Published: November 2022

Abstract
Water storage tanks are widely used in the nuclear industry for safety purposes, chemical water control and primary volume water management. Under seismic load, the sloshing of water inside the storage tank can cause a strong dynamic response, which may also change some mechanical characteristics of the tank, for example the center of mass. The water sloshing analysis is therefore an important part of the modal and seismic analysis of water storage tank. However, the water sloshing effect contains a complicated fluid-structure interaction but structural analysts cannot afford the time to spend on the calculation. It is important to build up a simplified equivalent model for the simulation.
EDF has been developing since 1989 its own in-house FEA code baptized code_aster, which is included in the SALOME-MECA mechanical package. In this paper two simplified fluid-structure interaction models are performed using SALOME-MECA. Firstly, an optimized Housner model is derived with formula and implemented with 3D finite model. Additionally, another 3D finite element model using added mass model assembled in code_aster, which can simulate more precisely the fluid-structure interaction of the water storage tank.
Based on the theory of these two models, their natural modes results and seismic response results are carefully compared and analyzed. The results indicate Housner model can provide a precise value of the water sloshing frequency and its displacement value is not accurate enough due to the impulse mass and convection mass simplification. However, the added mass model in SALOME-MECA can provide a more accurate and reasonable result on vibration mode calculation, which may provide an interesting alternative method for seismic verification for the nuclear refueling water tank.
Dynamic Wetting Behavior During Droplet Impacting on a Superheated Surface
Authors: Chen Mingjing , Chen Deqi , Liu HanZhou 
Published: November 2022

Abstract
During the reflooding process after a loss-of-coolant-accident (LOCA) scenario, the highly superheated fuel surface of the reactor core will undergo a quenching process. The temperature of the fuel element surface drops sharply during the quenching process, because large droplets can pass through the vapor film to re-wet the wall surface, and the boiling mechanism is transformed from film boiling to nucleate boiling, the boiling heat transfer perfomance is greatly improved. But the transition of the boiling mechanism is closely related to the initial velocity of the droplet, the size of the droplet, and the wall temperature. At the same time, after the droplets hit the wall surface, dynamic wetting behaviors such as deposition, vibration, rebound, or splashing will occur. Unfortunatly, the mechanism of droplet wetting behavior on boiling heat transfer is not clear, so it is necessary to carry out a fundamental research. In this study, the droplet size, initial droplet velocity, substrate’s temperature and surface tilt angle were precisely controlled. The images were captured by a high-speed camera to obtain a dynamic wetting image of the droplets impacting the wall from room temperature to 400°C. Finally, based on the image analysis method, a simple analysis of the three-phase contact line movement and the droplet dynamic wetting behavoir is performed.
Phase Distribution Characteristics of Bubbly to Slug Flow Transition Region in a 10mm ID Vertical Round Pipe
Authors: Xu Wangtao , He Qingche , Huang Tao , Ding Shuhua , Yan Meiyue , Zhang Luteng , Pan Liangming 
Published: November 2022

Abstract
As a common heat transfer structure, the vertical round pipe is widely used in nuclear energy systems. The phase distribution characteristics of two-phase flow in a vertical round pipe is an important issue. For two-phase flow, combined with the local typical flow phenomenon, the analysis of the phase distribution characteristics such as the void fraction and the interfacial area concentration is helpful to research the two-fluid model and interfacial area transport equation. Based on the completed flow pattern experiment, an air-water two-phase flow characteristic experiment is carried out for the visualized experimental section of 10 mm ID vertical round pipe. In addition, the 5 × 7 total 35 measuring points of bubbly to slug flow transition region are selected, including 5 superficial liquid velocities (0.5, 1.0, 1.5, 2.0, 2.5 m/s) and 7 superficial gas velocities (0.08, 0.11, 0.13, 0.17, 0.21, 0.28, 0.35 m/s). High-speed camera is used to capture the flow patterns of bubbly to slug flow transition region. Four-sensor conductivity probe is used to measure the interfacial parameters such as void fraction and interfacial area concentration. Thus, the phase distribution characteristics of bubbly to slug flow transition region in the round pipe is analyzed. The obtained two-phase flow parameter data can provide data support for the development of two-fluid model and interfacial area transport equation.
Investigation of Design Models for Printed Circuit Heat Exchangers With Straight Semicircular Channels
Authors: Mao Junjun , Zhao Gang , Yang Xiaoyong 
Published: November 2022

Abstract
Printed circuit heat exchangers (PCHEs) are a promising technology for the closed Brayton cycle of nuclear energy due to their high compactness, capability to withstand high pressure and large temperature difference, and high heat transfer efficiency. The design calculation of PCHEs has a significant impact on the cycle performance. This paper proposes a theoretical model for predicting the thermal-hydraulic performance of semicircular straight-channel PCHEs. The irregular fins are equivalent to rectangular fins of equal height, making the model more simplified and accurate. The model is verified using experimental data from the literature. Based on this model, the heat transfer characteristics of composite PCHEs, where crossflow exists at the inlet and outlet, are investigated. The results indicate that the heat transfer performance of PCHEs can be more accurately predicted by using the length of the overlapping region of hot and cold channels as the reference length. Furthermore, the effects of geometric parameters on the heat transfer performance of semicircular straight-channel PCHEs are studied. The results show that the heat transfer performance of the semicircular straight-channel PCHEs decreases with the increase of the channel radius and the vertical interval, and increases with the increase of the horizontal interval. And the horizontal and vertical interval affect heat transfer performance slightly. This study provides a reference for the design and application of PCHEs in the helium closed Brayton cycle in the future.
Safety Analysis for Transportation of Activated Secondary Neutron Source Assembly
Authors: Cao Nanfeng , Yang Shouhai , Xie Enfei , Jing Shunping 
Published: November 2022

Abstract
With the improvement of economic requirements of nuclear power operation, China General Nuclear Corporation (CGN) will gradually replace the more expensive primary neutron source assembly with the activated secondary neutron source assembly in the first cycle for reactor start-up. After activation, the secondary neutron source assembly needs to be adapted to the simulated fuel assemblies by underwater operation, and then integrated into a dedicated secondary neutron source assembly transport cask and transported to the reactor core. Therefore, the safety analysis for transportation of activated secondary neutron source assembly is important. Based on the NAC storage transport cask, the Monte Carlo program and CFX software are used for safety analysis mainly including structural evaluation, thermal evaluation and shielding evaluation. The result of structural evaluation shows that the containment is not breached under any of the normal conditions of transport or the hypothetical accident conditions. The result of thermal evaluation shows that the decay heat released by the activated secondary neutron source assembly can be derived effectively by the transport cask and the temperature of each component cannot exceed its corresponding limit. The shielding calculation results also meet the shielding acceptance criteria. Thus, this work verifies that the transportation of activated secondary neutron source assembly meets the design requirements in the main aspects of nuclear safety and is feasible.
Cause Analysis and Treatment Measures for Corrosion of Warm-Up Pipe of Hualong no.1 Main Feed Water Pump System
Authors: Wu Zhiming , Ling Yixin , Liu Xing , Fei Dongdong , Hou Xiaoyu , Wen Xue 
Published: November 2022

Abstract
The Main Feedwater System of Fuqing nuclear power plant is composed of three electric pump sets arranged in parallel with 50% capacity to provide the required Feedwater for the secondary side of steam generator. Under normal conditions, two pumps are in operation, one is standby, and three pumps can be switched at will. When one of the two running electric feed water pump sets trips, the electric feed water pump set in standby state will start quickly. If the equipment has problems, it will seriously affect the safe and reliable operation of the unit.
After the leakage defect of the warm-up pump pipe downstream of the warm-up pump stop valve in train C of the main feed water pump system of Hualong unit was found, through the disassembly inspection, it was found that there were many surface corrosion pits in the downstream pipeline of the throttle orifice plate. In this paper, through the physical and chemical analysis of the pipeline composition, organization, non-metallic inclusions, surface quality, macro and micro morphology, etc, At the same time, combined with the operating conditions of the pipeline, analyze the causes of the defects, judge the defects caused by the medium vaporization caused by the excessive throttling of the throttle orifice, and give effective treatment measures. Through these relevant measures, the corrosion rate of the warm-up pump pipe is reduced and the operation reliability of the important equipment of the unit is improved.
Engineering Application and Research Progress of Low-Level Radioactive Waste Incineration Technology in China
Authors: Haoran Chu , Wei Xu , Bowen Zheng , Jiasheng Ruan 
Published: November 2022

Abstract
China Institute for Radiation Protection had independently developed a multi-purpose radioactive waste pyrolysis incineration technology and built 3 incineration facilities in China, which were mainly used to treat solid waste and waste oil from nuclear facilities such as NPPs. The composition of solid waste included paper, cloth, plastic, rubber, etc. In order to solve the problems in the early operation of incineration facilities, targeted improvements had been made in the aspects of equipment anti-corrosion capability, secondary waste generation and system safety. The improved facilities had treated a large amount of low-level waste and operated for more than 15 years. The stability and reliability of the incineration system were verified, and the advancement of pyrolysis incineration technology was proved.
Considering the current situation that the proportion of plastics in low-level waste is increasing, the process was optimized so that more plastics and resin can be incinerated. The optimized incineration system still showed good adaptability when the proportion of plastic in the waste composition over 60%. The research on miniaturization and mobile technology of incineration were continuously carried out for the small reactors or small nuclear facilities, so as to further improve the economy. Compared with the incineration facilities, the mobile incineration technology can reduce the floor space by more than 90% and cut the construction cost by more than 75% under the same capacity and meeting the emission requirements.
Modeling and Analysis of Two-Phase Distribution in Large-Scale Separate Heat Pipe
Authors: Kuang Yiwu , Jin Xin , Zhuan Rui , Xie Fang , Chen Kaidi , Han Fei , Chen Xiting , Wang Wen 
Published: November 2022

Abstract
Large-scale separate heat pipe is widely used in various applications, such as waste heat recovery, passive cooling of Spent Fuel Pool (SFP) and so on. The evaporator and condenser of the heat pipe can be installed in different places with long distance. Working fluid circulates in the heat pipe due to static pressure difference between the evaporator and condenser. The separate heat pipe is a high efficiency heat transfer device which can transport heat for a long distance without considerable losses. A numerical model of separate heat pipe is established and verified according to the experimental data. The model agrees reasonably well with experimental data. The Mean Absolute Error (MAE) is 7.22%. According to the model, vapor quality at the evaporator exit in the separate heat pipe is analyzed. It is found that, vapor quality varies with heat pipe filling ratio and thermal loads. At low filling ratios, fluid at evaporator exit is overheated. While at high ones, it is in two-phase state. Vapor quality at evaporator exit decreases with the filling ratio. When the vapor quality is bit lower than 1, separate heat pipe comes to its maximum heat transfer capacity. It is also found that there is a minimum condenser height to activate the large-scale separate heat pipe. When the downcomer is full of liquid, elevating the condenser can increase the heat pipe performance. But when the condenser is high enough, the downcomer will be partially liquid filled, elevation of condenser provides little benefit.
Nuclear Security Against Insider Smuggling: Review and Recommendations
Authors: Al-Khodire Hiba 
Published: November 2022

Abstract
Smuggling of nuclear material is one of the serious problems of nuclear security. Nuclear institutes and organizations put great efforts to protect these materials from dangerous hands by applying nuclear security policies, using monitoring tools, and developing access control systems. These techniques effectively prevent outsider threats but cannot provide high confidence withstanding insider threats.
History offers many examples of violent employees, who broke the rules and smuggled nuclear material, taking advantage of their knowledge, experience, and authority to access the nuclear facility systems and their vulnerability. Nuclear smuggling operations can be through one insider, multi-insiders, or by the incorporation with an outsider, this makes the facility’s security planning a challenge.
The analysis of smuggling crime shows that the motivations for performing insider threats vary. It includes but is not limited to financial gain, poor security culture, and revenge. Lessons learned from other insider crimes in non-nuclear fields are useful for developing nuclear facility plans. Thus, it is enormously important to continuously improve the security system, overcome instrumentation malfunctions, and strengthen the nuclear security culture in the facility. This paper provides a review and recommendations in the legal and technical framework to genuinely prevent insider nuclear threats against nuclear facilities.
Parametric Study of Throttling Window Shapes on Flow Force Characteristics in Main Feed Water Regulating Valve
Authors: Zhao Lei , Ru Qiang , Wu Jia-Yi , Qian Jin-Yuan 
Published: November 2022

Abstract
As the key device of the secondary circuit system, the main feed water regulating valve plays a significant role in controlling water level, whose reliability directly affects the safety of the nuclear power plant. During the use of the main feed water regulating valve, the movement and stress state of the valve core are affected by the hydraulic force and the hydraulic torque, which is related to the accuracy and stability of the valve. In this paper, a typical throttling window type of main feed water regulating valve was taken as the research object, the shape coefficient of the throttling window was defined, and the effect of the throttling window shape on the force characteristics of the valve core under variable opening was analyzed. The results show that as the valve opening increases, the direction of valve core axial hydraulic force changes. And with the increase of the shape coefficient of the throttling window, the maximum axial hydraulic force and the maximum lateral hydraulic force of valve core first decrease and then increase, and the maximum hydraulic torque increases gradually. This study can provide some references for the optimal design of similar main feed water regulating valves.
Numerical Simulation of Single-Layer Wire-Mesh Sensor
Authors: Zhang Hengwei , Xiao Yao 
Published: November 2022

Abstract
Wire Mesh Sensor (WMS) is an advanced two-phase flow measurement device, which has been widely applied in experimental research and industrial production. Previous research has shown that the measurement uncertainty of the WMS increases with the increase of the axial distance between the two layers of the electrode wires. Moreover, the structure of dual layers of wire mesh limits the application of WMS in complex flow channels. For example, the structure of the rod bundle channel needs to be carefully designed to install WMS. Thus, a novel WMS with only a single layer of electrode mesh was proposed. The transmitter wires and the receiver wires were intertwined with each other to increase the stability. The signal response of the single-layer WMS was investigated by the potential field simulation. For the bubble with a diameter of two times the mesh size passing through the WMS, the signal response of the single-layer WMS is closer to the ideal signal compared with the two-layer WMS. For the bubble with a diameter of one time the mesh size, the single-layer WMS overestimates the void fraction. Near the intersection, the distance between the receiver and the transmitter is small, more currents would be received by the per length receiver wire. When the bubble contacts the single-layer WMS, it first wraps the highly sensitive wires, which leads to the void fraction being overestimated.
Research on Liability Insurance for Radioactive Sources and Suggestions to China
Authors: Li Xiao Ding , Gong Yu , Li Guang Hui , Bai Gang , Zheng Jie Ying , Qu Yun Huan , Li Zhe 
Published: November 2022

Abstract
Radioactive source liability insurance refers to the form of insurance in which the insurer shall be liable for compensation according to law when the insured uses radioactive sources to cause radiation damage or environmental pollution to a third party and is claimed for compensation. This paper deeply investigates the practical experience of various countries in the field of radioactive source liability insurance; this type of insurance has a lot of mature application experience in Europe & US, the two major nuclear damage compensation convention systems have not yet stipulated the liability insurance of radioactive sources, European countries have a strong awareness of applying for radioactive source liability insurance. Then it makes a detailed analysis of the current laws and regulations in the field of nuclear technology utilization and the current situation of liability insurance in the field of radioactive sources in China. At present, there is no special law on liability insurance in the field of radioactive sources, there are still several problems, including the weak awareness of risk prevention of enterprises using nuclear technology, the lack of social responsibility of some enterprises and their unwillingness to buy insurance.
In order to solve the above problems, the following four recommendations are made. First, improve the legal and policy system of radioactive source liability insurance in China. Second, pilot the implementation of radioactive source compulsory liability insurance in high-risk industries or enterprises. Third, promote personalized insurance rates and insurance limits. Fourth, carry out health damage insurance for employees in the field of radioactive sources at an appropriate time. Finally, it provides a strong and valuable reference for the future radioactive source liability insurance in China.