bluuebunny/crossref_metadata_2025_split huggingface dataset:
Cause Analysis and Influence Evaluation of Cracks in Thick Slab Construction
Authors: Guixiang Yi , Liang Li 
Published: October 2021

Abstract
Cracks were found in the slab of a nuclear power plant when the formwork was removed. By means of ultrasonic testing, water storage test and crack width detection, the fracture distribution, depth and width characteristics are determined. On this basis, the numerical simulation analysis of hydration heat in the maintenance process is carried out to simulate the possible crack generation and distribution of the floor slab in the maintenance process. The simulation result shows that cracks under the combined action of temperature and shrinkage are consistent with the cracking characteristics of floor slab. Through numerical simulation of the cracked floor and intact floor, the change of the out of plane bearing capacity (bending and shear) of the cracked floor is compared and analyzed. In the model, considering the reduction of the bearing capacity at the crack section and the change of load transfer effect at the crack interface, the concrete model is disconnected according to the known opening situation, and the contact relationship is set on the crack interface to simulate the crack. Through comparative analysis, the mechanical properties of the cracked floor are evaluated, and the refined numerical simulation method of the working components with cracks is proposed. This paper can provide guidance for the cause analysis and influence evaluation of similar thick plate cracking phenomenon.
Mixing Process of Two Component Gases by Natural Convection and Molecular Diffusion
Authors: Ube Takeaki , Takeda Tetsuaki 
Published: October 2021

Abstract
A depressurization accident involving the rupture of the primary cooling pipe of the Gas Turbine High Temperature Reactor 300 cogeneration (GTHTR300C), which is a very-high-temperature reactor, is a design-based accident. When the primary pipe connected horizontally to the side of the reactor pressure vessel of GTHTR300C ruptures, molecular diffusion and local natural convection facilitate gas mixing, in addition to air ingress by counter flow. Furthermore, it is expected that a natural circulation flow around the furnace will suddenly occur. To improve the safety of GTHTR300C, an experiment was conducted using an experimental apparatus simulating the flow path configuration of GTHTR300C to investigate the mixing process of a two-component gas of helium and air. The experimental apparatus consisted of a coaxial double cylinder and a coaxial horizontal double pipe. Ball valves were connected to a horizontal inner pipe and outer pipe, and the valves were opened to simulate damage to the main pipe. As a result, it was confirmed that a stable air and helium density stratification formed in the experimental apparatus, and then a natural circulation flow was generated around the inside of the reactor.
Research on Earthquake Acceleration Alarm of Nuclear Power Plant
Authors: Liang Li , Rong Pan , Guopeng Ren , Xiuyun Zhu 
Published: October 2021

Abstract
Almost all nuclear power plants in the world are equipped with seismic instrument system, especially the third generation nuclear power plants in China. When the ground motion measured by four time history accelerometers of containment foundation exceeds the preset threshold, the automatic shutdown trigger signal will be generated. However, from the seismic acceleration characteristics, isolated and prominent single high frequency will be generated the acceleration peak, which has no decisive effect on the seismic response, may cause false alarm, which has a certain impact on the smooth operation of nuclear power plant. According to the principle of three elements of ground motion, this paper puts forward a method that first selects the filtering frequency band which accords with the structural characteristics of nuclear power plants, then synthesizes the three axial acceleration time history, and finally selects the appropriate acceleration peak value for threshold alarm. The results show that the seismic acceleration results obtained by this method can well represent the actual magnitude of acceleration, and can solve the problem of false alarm due to the randomness of single seismic wave, and can be used for automatic reactor shutdown trigger signal of seismic acceleration.
Effects of Non-Condensable Gas on Characteristics of Natural Circulation Flow of Isolation Condenser
Authors: Takada Tetsuya , Yamamoto Yasunori , Ono Kosuke 
Published: October 2021

Abstract
An isolation condenser (IC) is a passive core cooling system in boiling water reactors. The cooling performance of IC is deteriorated when hydrogen generated in the core flows into the IC pipes. In this study, we conducted high pressure experiments using natural circulation loop with non-condensable gas injection, where helium was used to simulate hydrogen effect on the IC. The reaching distance of steam in the heat transfer tube was estimated by observing the region where nucleate boiling occurred on the outer surface of the heat transfer tube, and the heat transfer coefficient was estimated. The heat transfer coefficient hardly changed when helium was injected to the loop that indicates injected helium was not accumulated in the heat transfer tube. The system pressure at quasi-steady state increased with increasing amount of the injected helium. Since the differential pressure at the down comer section increased by helium injection, the injected helium may be accumulated in the section, leading to increment of the system pressure.
Recent Activities and New Challenges for the EUR Organization
Authors: Sorel Vincent 
Published: October 2021

Abstract
Since 1991, the European Utility Requirements (EUR) organization has been developing and promoting harmonized technical specifications for the new mid- and large-size LWR designs to be proposed by the vendors in Europe. The EUR Organization issues and, from time to time, modifies and updates a report entitled “the EUR Document”. It consists of a comprehensive set of requirements covering the whole Nuclear Power Plant (NPP) and encompasses all aspects (safety, performance, competitiveness) and all parts of a NPP (nuclear island and conventional island). The main EUR Organization’s results obtained during the last years are in the three following fields: (1) The latest revision, Revision E, of the EUR Document was issued in December 2016. It includes revised safety requirements taking into account the most recent European and international standards issued by WENRA, IAEA and IEC after the Fukushima-Daishi accident. (2) The assessment of new designs. The presentation briefly recalls the EUR design assessment objectives and process and the status of the different EUR assessment projects. (3) Preparation of evolution of the EUR Document towards integration of requirements for Small Modular Light Water Reactors (SMLWR). The paper concludes by an overview of the future activities of the EUR Organization.
Improved Wet Scavenging Schemes for Air Dispersion Modeling of Cs-137 in the Fukushima Accident
Authors: Zhuang Shuhan , Fang Sheng , Dong Xinwen 
Published: October 2021

Abstract
Wet scavenging process is critical for air dispersion modeling of Cs-137 in the Fukushima Daiichi Nuclear power plant (FDNPP) accident. Although intensively investigated, wet scavenging simulation is still subject to uncertainties caused by the biases in wet scavenging modeling and meteorological input. To reduce these uncertainties, the on-line coupled modeling feature of the Weather Research and Forecasting-Chemistry (WRF-Chem) model was utilized and both the in-cloud and below-cloud scavenging processes are considered. In this study, the in-cloud scheme Environ and below-cloud scheme Baklanov are combined with each other to form Environ-Bakla to simulate the wet deposition of Cs-137. The model is systematically compared with a previous WRF-Chem model with a single below-cloud scheme Baklanov, based on both the cumulative deposition and ambient concentration of Cs-137 based on the FDNPP accident observation. The results demonstrate that the in-cloud scavenging scheme substantially improves the cumulative deposition simulation in regions with light rain like Tochigi, Nakadori etc. With respect to the atmospheric concentration, the inclusion of in-cloud scavenging doesn’t necessarily improve the performances and the Environ-Bakla only shows fair performance under plume events with no rain or light rain.
Effect Analysis of Power Supply Topology on the Reliability of the Reactor Protection System
Authors: Zhang Haojing , Xiong Huasheng , Guo Chao , Li Duo , Huang Xiaojin 
Published: October 2021

Abstract
As a clean energy source, Low Temperature Reactor can be used for heating, desalination, industrial steam production, summer cooling, etc. It still has good application prospects at present. The reactor protection system is one of the important safety systems in Low Temperature Reactor, thus improving the reliability of the protection system can enhance the safety level and economic benefits of the reactors. The reliability of the protection system is directly related to its system structure, the quality of its components, and the power supply topology. The effect of the third factor is analyzed in this article.
In the aspect of system reliability analysis, the RBD method is widely used and has many public research results. Based on the block diagram of protection system, this article establishes the RBD that considers the operation failure. According to the foregoing RBD, the protection system reliability is calculated, the minimum cut sets that can cause the protection system failure are analyzed, the different importance measures of each device in the protection system is determined, and the occurrence probability of the minimum cut sets and the importance measures are sorted as a basis for subsequent optimization of power supply topology.
The Low Temperature Reactor protection system in this article is mainly composed of transducer, set-point comparison unit, logic coincidence unit, and safety trigger unit. The power supply of these units have many possible topological structures, which have different complexity and different reliability levels. Three improvement approaches are proposed for the power supply topological structure of transducer, set-point comparison unit and logic coincidence unit, the reliability of the optimized protection system is calculated and analyzed. The results show that by improving the power supply topology, the reliability of the protection system is improved without increasing the components’ quality and redundancy.
Experimental Study on Performance Improvement of HTR-10 Helium Purification System
Authors: Wang Fangfang , Wei Liqiang , Kang Tianyu , Hu Chuangguo , Xie Feng , Chen Xiaoming , Shi Lei 
Published: October 2021

Abstract
Based on the historical operation data analysis and experience feedback of the HTR-10 helium purification system, the performance improvement experiment scheme of the helium purification system equipment has been designed, and the simulation experiment bench with copper oxide bed and molecular sieve bed has been built. By sorting and analyzing the purification and regeneration experiment data of copper oxide particles (CuO-ZnO-Al2O3) and 5A molecular sieve, the following conclusions have been drawn: (1) The activity of copper oxide particles after initial activation is high and their performance decreases obviously after purification and regeneration. Electron microscope scanning shows that sintering deactivation phenomenon exists in the application of CuO, and its regeneration times and service life should be concerned; (2) The operation mode of molecular sieve bed in series is more favorable to the purification of impurity gas, and moderate high temperature and large flow sweep are beneficial to the regeneration and desorption of molecular sieve. The optimized process scheme can further improve the performance of HTR-10 helium purification equipment, and reduce the content of helium impurities in the primary circuit in future, which can provide a certain reference significance to achieve the ultra-high temperature operation of HTR-10.
Research on DTS Analysis Method for 1000MWe PWR NPP
Authors: Zhang Yubin 
Published: October 2021

Abstract
The design transients (DTS) are intended to be used to evaluate component stress analysis, which bound the plant operation over the design plant life, and will be as a reference document regarding plant usage factor surveillance. The conditions of DTS represent bounding cases for plant events that are expected to occur, or that may occur, during the plant lifetime. For different operating conditions, DTS analysis methods would be different.
This report describes the DTS analysis method to be used when designing the major reactor coolant system components of the 1000MWe PWR plant. Some of components will be considered: Reactor Pressure Vessel, Main Coolant Lines, Reactor Coolant Pumps, Pressurizer and Steam Generators. Generally, variations in fluid pressure, fluid temperature and flowrate are used to represent the parameters for DTS. This report describes the method of DTS analysis and shows the evolution process of relevant thermal hydraulic parameters.
High Flux Reactor Review and Reactivity Control Analysis
Authors: Lin Wang , Wei Xu , Fei Xie 
Published: October 2021

Abstract
For over 60 years, research reactors have provided the world with a versatile tool to test materials and promote irradiation research, as well as to produce radioisotopes for medical treatments. The High Flux Reactor (HFR), as a water moderated and cooled, beryllium-reflected reactor has awarded more attention in recent years. There is a wide range of designs and applications for HFRs that based on their own situation to meet research requirements. For the purpose of reducing the volume and mass of the reactor, as well as ensuring the safety operation, it is necessary to determine the most effective reactivity control scheme, and analyze the corresponding reactivity insertion accidents. This paper is going to investigate typical high flux reactors within the same type with HFETR, summarize general description and characteristics, the uses of the high flux reactor, and reactivity control mechanisms. In addition, the associated reactivity insertion accidents were presented and analyzed. The analysis result will provide some references to further design and construction of high flux reactor.
Experimental Study on Flow Pattern of 10 mm Vertical Pipe
Authors: Xu Wangtao , He Qingche , Yan Meiyue , Zhu Wen , Zhang Luteng , Wu Dan , Huang Tao , Ma Zaiyong , Sun Wan , Pan Liangming 
Published: October 2021

Abstract
As a common heat exchange structure, vertical pipe is widely used in neaclear reactors and other systems. Flow pattern and flow regime map are the basis of thermal-hydraulic calculation. In this paper, the experiment of air-water two-phase flow pattern was carried out based on the visual test section of 10 mm vertical pipe. Three impedance meters were used to measure the void fraction at three axial positions (L/D = 66, 148, and 229) along the vertical pipe. Compared with the flow pattern criterion proposed by Mishima and Ishii, six superficial liquid velocities are selected for design conditions, which are 0.5 m/s, 1.0 m/s, 1.5 m/s, 2.0 m/s, 2.5 m/s and 3.0 m/s, respectively. And the superficial gas velocities are ranging from 0 to 30 m/s. Considering the inlet and export effects of pipe flow, the void fraction parameters from L/D = 148 (#2 impedance meter) were applied for analysis. Based on the experimental data and random forest algorithm, the flow pattern was identified for each condition, and the flow regime map of 10 mm vertical pipe was established. It is concluded that the flow pattern transition point of air-water flow in the 10 mm vertical pipe was different from that of the Mishima-Ishii’s model, and the inner diameter of the pipe will affect the flow pattern formation and flow pattern transition.
Research on Dimensionless Analysis Method of Scale Effects for Molten Pool Experiments
Authors: Quan Fengyang , Li Wei , Zhao Zikun , Zeng Xiao , Guo Yong , Yuan Yidan 
Published: October 2021

Abstract
In a severe accident of light water reactors, the loss of coolant and heat sink can cause core degradation. The molten fuel pellet (UO2), molten cladding material (Zr-ZrO2), and molten structural materials (stainless steel) form molten corium, and finally relocate in the lower head of the reactor pressure vessel (RPV). In case of coolant loss, the corium molten pool form in the lower plenum which will threaten the integrity of RPV. To prevent the RPV failure, the ex-reactor vessel cooling (ERVC) system, which floods the RPV with cooling water, is applied to ensure the survivability of RPV and eventually achieve the goal of fission product retention. Therefore, the molten pool’s behavior is well studied by a large number of experiments, such as BALI, LIVE-2D, COPRA. Natural heat convection, which is related to the latent heat of fission products and the cooling of ex-vessel, will induce heat flux alone the RPV lower plenum. The peak value of heat flux is estimated at around 70°. However, there is still uncertainty to predict the Nuup/Nudn ratio especially when the condition of the prototype molten pool (1017) to be extrapolated. This research aims at analyzing the scale effects based on Buckinhum’s PI theory. The molten pool phenomenon related to dimensionless parameters is firstly calculated. After rigorous calculation, the influence of each dimensionless parameter is evaluated according to the behavior of the corium pool. Finally, the experimental results of BALI and COPRA are obtained, in order to evaluate the uncertainty between the chosen simulant from those two experiments.
A Preliminary Evaluation of the Computational Fluid Dynamics Capabilities in MOOSE
Authors: Weiss Abdullah G. , Abdoelatef M. Gomaa , Bani Ahmad Mohammad T. H. , Ahmed Karim , Kimber Mark L. 
Published: October 2021

Abstract
The Multiphysics Object Oriented Simulation Environment (MOOSE) is a code package that couples a variety of physics modules, allowing for highly accessible multiphysics simulations. The physics modules include a finite-element Navier-Stokes (N-S) module that is designed to solve laminar fluid dynamics problems. The usage of this module in multiple recent studies coupled with the growing interest in MOOSE for usage in non-LWR safety studies by the Nuclear Regulatory Commission (NRC) prompted the authors to investigate the computational fluid dynamics (CFD) capabilities of MOOSE. A 2D flow past a round cylinder scenario is simulated in the MOOSE environment to investigate the effectiveness of the N-S module. The simulation assumed an unsteady laminar flow with a Reynolds number of 200, where the unsteadiness in the vortex street is captured using a Euler-implicit time-integration scheme. To verify the results from MOOSE, a similar simulation is conducted using the well-utilized OpenFOAM finite volume code, and the velocity profiles of both transient simulations are compared. Grid and input sensitivity studies are conducted in MOOSE to quantify the numerical and input errors, respectively. The numerical and input errors are also visualized with contour plots in order to qualitatively understand the evolution of the errors across time and space. Results show that MOOSE struggles with capturing the instabilities in the flow, as it takes more than 20 seconds before a vortex street is observed. Excellent agreement in the velocities is seen between MOOSE and OpenFOAM upstream of the cylinder, but huge discrepancies can be seen where instabilities are expected around the cylinder and upstream of it. The discrepancies become more severe as the transient simulations progress in time. The numerical error is most prominent in the wake of the cylinder and around the boundaries of the channel. The input errors are also prominent in the wake but are most prominent around the cylinder walls. The results demonstrate the difficulties faced by the N-S module in MOOSE when it comes to simulating unsteady flow and prompts questions about the effectiveness of finite element CFD solvers.
Numerical Calculations of the Effective Thermal Conductivity of the Dispersion Fuel Sphere With the Internal Heat Sources
Authors: Ziping Liu , Yu Ji , Han Zhang , Jun Sun 
Published: October 2021

Abstract
The effective thermal conductivity of dispersion fuel sphere is an important parameter for its’ temperature prediction. Whether the existing effective thermal conductivity models which does not consider the impact of inner heat source can predict the average temperature of heat generating dispersion fuel sphere is unknown. This study established the three-dimensional heat conduction model to calculate the temperature distribution of dispersion fuel sphere with inner heat source, then the effective thermal conductivity for average temperature prediction was calculated by conduction differential equation. The fuel particles are set to be uniformed with uniform inner heat source, fuel particles are defined uniformly distributed in probability in the fuel sphere. Through the analyses, the existing models overestimate the effective thermal conductivity of dispersion fuel sphere with inner heat source, and then underpredict the average temperature of dispersion fuel sphere. The distribution of inner heat source will affect the effective thermal conductivity. The values of inner heat source will not affect the effective thermal conductivity. This paper revealed the importance of inner heat source on effective thermal conductivity, and lay the foundation for further study.
Research on Optimization and Verification Method of Sensor Arrangement in the Chemical and Volume Control System
Authors: Zhou Gui , Wang Hang , Peng Minjun 
Published: October 2021

Abstract
In order to avoid the nuclear accidents during the operation of nuclear power plants, it is necessary to always monitor the status of relevant facilities and equipment. The premise of condition monitoring is that the sensor can provide sufficient and accurate operating parameters. Therefore, the sensor arrangement must be rationalized.
As one of the nuclear auxiliary systems, the chemical and volume control system plays an important role in ensuring the safe operation of nuclear power plants. There are plenty of sensor measuring points arranged in the chemical and volume control system. These sensors are not only for detecting faults, but also for running and controlling services.
Particle swarm algorithm has many applications in solving the problem of sensor layout optimization but the disadvantage of the basic particle swarm optimization algorithm is that the parameters are fixed, the particles are single, and it is easy to fall into the local optimization. In this paper, the basic particle swarm optimization algorithm is improved by Non-linearly adjusting inertia weight factor, asynchronously changing learning factor, and variating particle. The improved particle swarm optimization algorithm is used to optimize the sensor placement.
The numerical analysis verified that a smaller number of sensors can meet the fault detection requirements of the chemical and volume control system in this paper, and Experiments have proved that the improved particle swarm algorithm can improve the basic particle swarm algorithm, which is easy to fall into the shortcomings of local optimization and single particles. This method has good applicability, and could be also used to optimize other systems with sufficient parameters and consistent objective function.