bluuebunny/crossref_metadata_2025_split huggingface dataset:
Design of the Protection Device Withstand External Explosion Shock Wave for the Nuclear Power Plant
Authors: Zhang Mingqian , Li Zhenguang , Wang Qingli 
Published: November 2022

Abstract
The design of the nuclear power plant shall meet the general objective to ensure that the safety functions of the systems and components are not affected by any external explosion hazard. A protection device must be sufficiently designed to protect the target equipment from the external explosion load. The shock wave propagation and attenuation was simulated dynamically for a new protection device withstand external explosion by use of the computational fluid dynamics software FLUENT. The method used in the simulation was firstly validated by comparing the simulation and the results of the shock tube test. Furthermore, the numerical model of the shock wave propagation and attenuation through the protection device was developed by this method, and the pressure time history behind the protection device was obtained, and then the protective effect of more layers of the protection device and the attenuation effect of the protection device preventing the lower amplitude of the incident shock wave were taken into account. It was found that the protection device could provide an effective limitation to the external explosion, the protective effect had not been strengthened obviously with more layers of the same protection device, and the attenuation effect would decrease with the lower amplitude of the incident shock wave. This analysis practice predicts virtually the physical phenomena, and can be taken as a guidance for the protection device design development and improve our understanding of external explosion shock wave phenomena.
Control Strategies Evolution of Nuclear Plant: From Obninsk to HTR-PM
Authors: Cheng Zhonghua , Dong Zhe , Li Bowen 
Published: November 2022

Abstract
Since the first commercial nuclear plant Obninsk began to operate, there have been dozens of types of nuclear plants studied, designed, or constructed nowadays driven by the key issue: how to enhance the economic competitiveness of nuclear plants while ensuring safety. Research on the key issue stimulated the evolution of nuclear plants, which simultaneously brought an array of various control problems for each generation and ultimately attributed to the evolution of control strategies of nuclear plants. In this paper, the main development route of nuclear plants is presented with a concise introduction to the representative design of nuclear plants for each generation. Meanwhile, the corresponding control problems brought by the design and/or the safety requirement document, national and international requirements, and so forth, were analyzed with emphasis on the Gen IV multi-modular high-temperature gas-cooled reactor (MHTGR). On account of the requirement to cooperate with other renewable energies to achieve the “Emission peak, Carbon neutrality” through the improvement of operation flexibility of NPP, most SMR plants work in the reactor following mode which requires more R&D efforts on the coordination control and load-following control to enhance economic advantages of SMR plants, etc.
Development and Verification of Containment Heat Sink Module in Severe Accident Code
Authors: Yang Zhi , Zhou Peng , Chen Lin , Cheng Yixuan , Zhang Hao , Yang Yanhua , Mao Yichen 
Published: November 2022

Abstract
As an important part of COSINE software package, cosSA code is used for safety analysis of severe accidents in nuclear power plants. The containment heat sink can transfer the heat from the containment to the environment and thus decrease the risk of the containment failure, especially when the primary loop is broken. The containment heat sink module provides heat transfer calculation from gas space and water pool to the heat sink, so as to accurately predict the temperature of containment heat structure and gas temperature and pressure in the compartments. In this paper, the development of the containment heat sink module in cosSA code is introduced, including lumped parameter heat sink, distributed parameter heat sink, heat sink surface heat transfer, and inner heat conduction. In order to verify the current containment heat sink module, a series of tests are designed based on a simple case with three compartments. In general, the test results are in good agreement with the similar software for the totally submerged and totally uncovered heat sink, indicating the effectiveness and accuracy of the current containment heat sink module. For the partial submerged distributed parameter heat sink, the transfer to the gas space and water pool are calculated simultaneously. The axial division method can alleviate the extra heat transfer effect from gas space to water pool.
Development and Verification of Multigroup Cross Section Generation Module GroupXS in AXSP
Authors: Hu Kui , Ma Xubo , Ma Xuan , Zhang Chen , Jia Guanqun , Chen Yixue 
Published: November 2022

Abstract
It is important for nuclear reactor physics calculation to generate accurate multigroup cross section libraries by using evaluation nuclear data libraries. A new evaluated nuclear data library processing code, named as AXSP (Advanced Cross Section Processing Code) was developed with independent intellectual property rights. The GroupXS module was developed with the ability of generating multi-group cross-sections based on the Bondarenko method which is always a favorable approximation for the fast spectrum reactor, and the method was also used in GROUPR of NJOY. To verify the accuracy of the GroupXS module, the GROUPR module of NJOY was used, and we corrected three numerical verification bugs of GROUPR. The bugs were described in this study. By comparing the calculation results with GROUPR of NJOY2016, the relative error of multigroup flux, multigroup cross sections of various reaction types are less than 0.01%, and the relative errors of transfer matrices of various reaction types are basically less than 0.1%. All the neutron reaction types of all nuclides in ENDF/B-VII.1, ENDF/B-VIII.0 and CENDL-3.2 can be processed by GroupXS, and the multigroup cross sections generated by GroupXS of AXSP have a good agreement with that generated by GROUPR of NJOY2016.
A Space Reactor Control System Design and Simulation Platform Development and Verification
Authors: Liu Pinchao , Sun Peiwei 
Published: November 2022

Abstract
A liquid metal cooled reactor coupled by mixed gas of helium and xenon closed Brayton cycle is designed for space exploration purpose. The design and dynamics of the space reactor differ from those of traditional reactors. Therefore, it is necessary to develop a control system design and simulation platform to investigate its dynamic behavior, design and evaluate the control system. Two system codes are adopted and coupled to develop the platform. The reactor core system code is used to calculate the reactor kinetics and thermal-hydraulics of the coolant. The energy conversion system code is to compute the Brayton cycle and the heat exchanger between the primary and secondary loops. 3KeyMaster simulation platform is used to load the two system codes by an interface program through dynamic link library. The data exchange and time synchronization between 3KeyMaster simulation platform and two system code are performed to realize the coupling of two system code. By introducing typical disturbances, the calculation results are compared with those from the system codes to verify the interface program. The input and output parameters can be visualized by model components in 3KeyMaster simulation platform to analyze the dynamic behaviors, which is useful for the subsequent control system design.
The Location of Spent Fuel Pins in Heavy Water Research Reactor in the Spent Fuel Pool
Authors: Lu Jin , Nie Peng , Ren Ren , Guo Yaping , Wang Qi , Zhang Xinwang , Zhang Lijun 
Published: November 2022

Abstract
Heavy Water Research Reactor was the first reactor in China. It adopted metallic uranium as fuel and aluminum alloy as cladding in the beginning. During the refueling procedure many years ago, several fuel pins dropped at the bottom of the spent fuel pool. The amount and the location of the fuel pins were not recorded. Before decommissioning, it is important to locate and take back the fuel pins. The fuel pin is φ40 mm × 100 mm and the pool is 17.8 m × 5.6 m. It is like dredging for a needle in the sea. What’s worse, it is not clear to find the fuel pins even with a waterproof camera for the bad water quality. Besides, radiation tubes, neutron detectors, and broken control rod were also stored in the pool.After reviewing on the localization method of radioactive source at home and abroad, it is determined to use γ dose rate meter and γ spectrometer in this case. Shields in different width and thickness were calculated with MNCP code. The optimization was 5cm in thickness and 5 cm width lead shield with a hole in the center. The waterproof camera was tied with Gama detector and finally 14 fuel pins were located and taken back safely.
Simulation Research on Real-Time Offsite Dose Radiation of Nuclear Power Plant
Authors: Ou Pingwen , Wang Zhenghui , He Dongyu , Jiang Pingting 
Published: November 2022

Abstract
With the rapid development of nuclear power, its safety has attracted more and more attention. In the event of a nuclear accident, once the leakage of radioactive material occurs, it will pose a huge threat to the surrounding environment and the economy and safety of human. In this case, the rapid and accurate evaluation of pollutant dispersion and radiation dose of offsite consequences can guide the decision of nuclear emergency, including the public emergency response and rescue operations. It can also provide the best solution of routes for public evacuation and rescue entrance. The limitation of radioactive accident consequences is essential to protect property security and public safety.
In this paper, a real-time evaluation program for offsite dose for nuclear emergency system is developed. The selected for atmospheric dispersion model is the segmented Gaussian model (SGM). Diffusion, advection, plume rise, dry deposition, wet deposition, decay, etc. are considered and FORTRAN language is used for the development.
For code-to-code comparison, C3X, the French general offsite dose evaluation program, is chosen as the reference. In the release of changing weather conditions, the simulation results are consistent with C3X, which demonstrates the validity and reliability of program. This paper makes also a nuclear emergency simulation for the changing weather conditions. The results show that the program can precisely calculate the plume nuclides concentration, ground concentration and dose in different time and space. It can also accurately divide the dose area for public emergency guide and rescue operations.
Modal Simulation and Verification on Stator End Winding of Large Nuclear Power Turbine Generator
Authors: Yang Xike , Zhang Jian 
Published: November 2022

Abstract
The 1300 MW nuclear power turbine generator is a four-pole generator with a rated speed of 1500 r/min. Due to its large capacity and low speed, the stator core and stator end winding bear huge electromagnetic torque. At the same time, because of the large heat generated by the stator winding during operation, the stator end winding is subjected to a large thermal expansion force. All these loads may have a serious impact on the stator end winding. In this paper, the three dimensional parametric finite element method and modal optimization algorithm are used to obtain the finite element model of each component of the stator end winding. Finally, the modal natural frequency of the overall end winding structure is obtained, so that the four-lobe modal natural frequency avoids the excitation frequency of 100 Hz, and the accuracy of the simulation mode is verified by modal test. The stator end winding modal simulation provides a reference for the stator end winding structure design of the large nuclear power turbine generator, ensures the good dynamic characteristics of the end winding structure, which provides a guarantee for the safe and stable operation of the stator end winding.
Modelica Based Economic Analysis of Nuclear Fuel Cycle
Authors: Liu Jie , Wang Xiang 
Published: November 2022

Abstract
The nuclear fuel cycle with spent fuel reprocessing process is called closed cycle. The reprocessing process of the fuel cycle recovers the fission elements uranium, plutonium and their isotopes. The cost of the nuclear fuel cycle includes front-end and back-end. The front-end segments mainly include mining, conversion, concentration, fuel element fabrication, etc. The back-end cost includes storage, processing, disposal and so on. Modelica language is a physical process description language based on objects and equations. Modelica language can be used to realize the visual modeling of various causal or non-causal processes. The main content of this research is to model different nuclear fuel cycles, including summarizing the simplified model of closed fuel cycle system, economic considerations and system working principle, the current economic analysis and research status of closed fuel cycle, combing the existing models, the use and application of OpenModelica language, especially the visual modeling method of OpenModelica, to achieve economic analysis of different nuclear fuel cycles.
A Weakly Supervised Time Series Analysis Framework for Anomaly Detection in Nuclear Power Plants
Authors: Dong Feiyan , Chen Shi , Demachi Kazuyuki , Yoshikawa Masanori , Seki Akiyuki , Takaya Shigeru 
Published: November 2022

Abstract
Condition monitoring is essential to the management and maintenance of Nuclear Power Plants (NPPs), as anomalies in the condition of components can affect the normal operation state of the entire plant. Therefore, timely and automatic detection of anomalies plays an important role and is in high demand. At present, deep learning is widely used for anomaly detection. Nevertheless, anomalies are difficult to define, sparsely occurring, and are accompanied by variable noise labels, which poses challenges to detection. Moreover, the problems such as loss of temporal features and gradient vanishing that exist in general deep learning models when dealing with time series data also increase the difficulty of anomaly detection. In response to these problems, a weakly supervised time series analysis framework for anomaly detection in NPPs is proposed, constituted of weakly supervised learning (WSL) and attention mechanism. The validation of the proposed framework was performed on the High Temperature Engineering Test Reactor (HTTR) anomaly cases dataset using the analytical code “ACCORD”, which distributed anomalies independently across multiple instruments and were recorded from the responding sensors to each anomaly. At this stage, 3 classes of anomalies were used as input data for the validation experiments. The experimental results demonstrate the effectiveness and feasibility of the proposed framework on anomaly tasks for the condition monitoring of NPPs.
Developing Intelligent Equipment Monitoring System of Nuclear Power Plants to Improve Plant Performance
Authors: Yu Jianming , Cai Yibiao 
Published: November 2022

Abstract
The active development of nuclear power is an important strategic choice to achieve carbon peak and neutrality goals. It is an essential study to realize the economic operation of nuclear plants on the premise of ensuring the safety and reliability of nuclear power plants. Intelligent monitoring and maintenance technology is an important development tendency that can effectively improve the operation safety and maintenance economy of nuclear power plant. In this article, an intelligent monitoring and maintenance system of AP1000 significant equipment in Sanmen Nuclear Power Plant is developed based on the equipment operation mechanism and failure mode analysis by using advanced pattern recognition technologies including intelligent sensing, big data analysis, multidimensional spatio-temporal correlation condition monitoring and fault diagnosis technologies. It can realize online real-time equipment performance monitoring, automatic trend analysis, automatic health condition evaluation, early anomaly identification and automatic fault diagnosis of equipment in nuclear power plant and promote the transformation of maintenance mode from time-based maintenance to condition-based maintenance. The intelligent monitoring and maintenance system has been applied in various mechanical and electrical equipment of Sanmen Nuclear Power Plant such as reactor coolant pump, main steam turbine, main transformer, heat exchanger and achieved certain technical achievements and economic benefits. The construction of intelligent monitoring system has certain demonstration and popularization.
Estimation of Decommissioning Radioactive Wastes Quantity for Nuclear Power Plant
Authors: Li Senquan 
Published: November 2022

Abstract
Most of the radioactive waste generated during the decommissioning of nuclear power plants are relatively low in radioactivity, except for the reactor pressure vessel, reactor internals and primary shielding walls, which have high levels of radioactivity. However, the treatment and disposal costs of decommissioning waste account for a high proportion of the total decommissioning cost. In order to minimize the decommissioning waste and to plan the decommissioning costs, it is necessary to study and estimate the total quantity of decommissioning radioactive waste in the design phase. This paper presents the statistical method for estimation of decommissioning radioactive waste quantity, gives the complete steps for estimating decommissioning radioactive waste quantity, and determines the waste quantity estimation principle for each type of decommissioning radioactive waste. Finally, the estimation task based on a Chinese typical million kilowatt nuclear power station single unit is reasonably completed. It is estimated that the quantity of decommissioning radioactive waste generated by this nuclear power station single unit is about 12000m3, which is similar to that of the same type-unit in the world. This proves the rationality of the method for estimating decommissioning radioactive waste quantity proposed in this paper.
Experimental Study on Measurement Characteristics of Different Type Flow Meters in Pb-Bi Environment
Authors: Liu Xin , Yun Xiangyu , Liang Junming , Zhang HuiYong 
Published: November 2022

Abstract
Lead bismuth eutectic is one of the most promising materials for coolant of fourth-generation reactor due to its outstanding heat transfer performance and chemical stability. However, the density and viscosity of lead bismuth eutectic is quite different from water, leading to a difficulty for flow measurement because of the high temperature operating environment and its strongly corrosive. Therefore, there is currently not a reliable flowmeter for measuring the volume flow of the lead bismuth eutectic. Electromagnetic flow meter has the advantage for measuring lead bismuth eutectic flow, due to the non-contact measurement, simple structure and high temperature resistance, so it has become one of the main selection flowmeters for lead bismuth eutectic flow measurement. And venturi flowmeter is one of the most widely used flowmeter types in water media. Therefore, the measurement characteristics of the electromagnetic flow meter and the venturi flowmeter in liquid lead bismuth eutectic environment had been studied.
In this context, two type flow meters had been calibrated by the volume method. The influence of the temperature and other factors had been studied on these two flowmeters.
The experimental results showed temperature affected the electromagnetic flowmeter measurement results, but it had no effect on venturi flowmeter. And the flow coefficient of electromagnetic flowmeter has a linear relationship with the temperature. And the relationship between temperature and flow coefficient of electromagnetic flowmeter was fitted by the calibrated data. Also, the flow coefficient of venturi flowmeter had been corrected too.
Also, the influence of discharging and refilling the LBE had been studied. And the results showed there is no influence on two type flowmeters. So, there is no necessary to calibrate the flowmeter repeatedly and the experimental process of LBE can simplify.
Radiation Shielding Towards Commonly Available Objects
Authors: Zamani Ahmad Muhammad Amir Safwan , Sazali Muhammad Arif , Ripin Azuhar 
Published: November 2022

Abstract
Radiation exposure is an essential measure of social welfare. Every day the public was exposed to radiation, whether they were aware of it or not. Public people can only receive no more than 1mSv of radiation in a year based on Basic Safety Radiation Protection Regulations 2010 under Atomic Energy Licensing. If it exceeds the value, it may risk people’s health and well-being. People are constantly exposed to radiation sources that exist naturally in everyday life. Without any action taken, the radioactive sources might penetrate people’s bodies. A shield is needed to avoid these unnecessary radiations. However, radiation shield are not comfortable to use daily. Thus, the main objective of this experiment is to observe the effectiveness of a commonly available object as a radiation shield. These objects can be found easily daily and used as a radiation shield. Gamma sources that are used in this experiment is 241Am and 133Ba. Solid-state detectors are used to measure the counts. The experiment shows positive feedback from glass, dark glass, brick, clay, and battery. They can shield gamma rays effectively. The main element of these objects is observed to see how the elements affected the objects’ ability to shield gamma. Carbon, oxygen, sulphur, hydrogen, and zinc can be used as the main elements to produce radiation shielding. The supervisor continuously monitored the execution and the safety during the experiment.
Numerical Simulation of Reactor Neutron Noise Based on Discrete Ordinate Finite Element Method
Authors: Yuan Baoxin , Luo Xin , Zeng Herong , Huang Huan , Zheng Jie , Yang Wankui , Zhang Songbao , Zhong Bin , Wei Junxia , Ying Yangjun 
Published: November 2022

Abstract
At present, the technology of reactor neutron noise is developing rapidly, which can not only be used to measure some intrinsic parameters of the reactor core, but also be used for core fault diagnosis. The theoretical solution of reactor neutron noise is of great significance for providing information on the directivity of core faults, and its analytical solution is extremely difficult or even impossible. Therefore, it is necessary to study the numerical simulation method of reactor neutron noise. In this work, the governing equations of the neutron noise problem are firstly established based on the discrete ordinate finite element theory, the equations take the reactor steady-state parameters and the core disturbance cross-section as input parameters. Secondly, the frequency domain solver FrequencyDomain_SNSolver of neutron noise is developed, which is capable of acquiring steady-state parameters, reading disturbance cross-section and calculating neutron noise. Finally, for a calculation case of an experimental reactor core, the neutron noise distribution and spectrum under perturbation are obtained by using the developed FrequencyDomain_SNSolver. This work can provide some reference and technical support for related research.