bluuebunny/crossref_metadata_2025_split huggingface dataset:
Creep Properties of FeCrAl Alloy at High Temperature Under Neutron Irradiation
Authors: Yao Huan , Ye Tianzhou , Wu Junmei , Wu Yingwei , Yin Chunyu , Chen Ping 
Published: November 2022

Abstract
Nuclear fuel cladding is subjected to neutron irradiation in a high-temperature stress environment, and the structural integrity of the cladding is very important for the safe operation of nuclear reactors. FeCrAl alloy has become a promising candidate cladding material for the accident tolerance fuel development in view of its excellent irradiation resistance and high temperature strength. This work aims to study the creep properties of FeCrAl alloy at high temperatures under neutron irradiation. Thermal and irradiation creep behavior in nanocrystalline FeCrAl samples is examined using molecular dynamics simulation method. And the effects of temperature, stress, irradiation dose rate on the creep rate and parameters of the creep constitutive equations are discussed. The results show that the thermal creep rate is greater than irradiation creep rate. The effect of temperature on the thermal creep stress exponent is relatively small at low stress, but is obvious when stress exceeds 0.8 GPa. The higher the temperature, the larger the thermal creep stress exponent. The irradiation creep rate increases almost linearly with the dose rate, that is, the exponent of dose rate for irradiation creep approach 1.0. Irradiation creep stress exponent fluctuates very little around 1.1 within the scope of the present research. Besides, higher temperature accelerates the linear increase of irradiation creep rate with dose rate, and the irradiation creep pre-factor becomes higher.
Research on Scenarios and Development Paths of China’s Commercial Closed Nuclear Fuel Cycle
Authors: Xiao Min , Xiao Xiaojun 
Published: November 2022

Abstract
China implements the established policy of closed nuclear fuel cycle for the sustainable development of nuclear power. However, there seems no feasible development plan and roadmap to initiate and deploy a commercial closed fuel cycle in China up to now.
The industrialization of the nuclear fuel cycle requires gradual and phased progresses. Since most of the operating nuclear power plants in China are PWR units, and China implements and promotes a commercial closed nuclear fuel cycle, how to initiate a closed nuclear fuel cycle from the mature commercial pressurized water reactors is an unavoidable reality.
Different from the implementation of closed nuclear fuel cycle reactors in countries such as France and Russia, the operating status and modes of PWRs in China are varied significantly. Most PWRs in China have implemented different plant modifications such as reactor power upgrading, core design and fuel management improvements with different new fuel types, different burnups and different cycle lengths, which have consumed certain degree of safety margins. These characteristics and differences bring challenges and difficulties to the implementation of a closed nuclear fuel cycle in China.
Based on international experiences and China’s situation, this paper discusses the necessity of initiating a closed nuclear fuel cycle from mature commercial nuclear power plants in China as the initial stage of the closed fuel cycle to lay the foundation for the future advanced nuclear fuel cycle, analyze and discuss the initiating mode of China’s commercial closed nuclear fuel cycle, review the nuclear fuel types to be utilized in the closed nuclear fuel cycle, and discuss the possible configuration and development path of China’s closed fuel cycle in the future.
Research on Signal Reconstruction Technology for Nuclear Power Plant Based on Self-Associative Kernel Regression
Authors: Miao Zhuang , Yu Fangxiaozhi , Xu Zhao 
Published: November 2022

Abstract
The instrument control system of the nuclear power plant is an important part of ensuring its safe and reliable operation. The sensor is the core component of the instrument control system to monitor the operation status of the nuclear power plant and trigger the operation of the instrument. Therefore, the availability of the sensor plays an important role of the safe operation of the nuclear power plant. The virtual sensor technology based on self-associative kernel regression can accurately determine the fault of the sensor signal, such as instrument drift, missing instrument values, etc. And it can replace the fault sensor with reliable and accurate values in a short period of time, ensuring normal operation of the system in a short period of time. In addition, the output value of the virtual sensor can also be used to calculate the measured value of the unmounted sensor part. This paper introduces a method to replace sensor values with right values when measured sensor values has something unusual. This paper only concerns about replacing sensor values and the paper assumes that it can find sensor fault correctly and timely. The paper selects the main steam system of the pressurized water reactor of the nuclear power plant as an example to verify the feasibility and reliability of the virtual sensor technology based on the self-associative kernel regression. The research results show that it is basically feasible to implement the virtual sensor technology by using the related technical solutions proposed in this paper.
Relay Chatter Analysis With Practical Application in Seismic PSA in Nuclear Power Station
Authors: Yang Jianfeng , Gu Xiaohui , Feng Bingchen , Wang Hangding 
Published: November 2022

Abstract
In the seismic PSA analysis of a nuclear power plant, the relay chatter analysis is a unique and important research content. An earthquake may cause inadvertent vibration of the relay contacts with an undesired output due to the low seismic ruggedness, and if the chatter time sustained more than 2 milliseconds, it will give an error signal. The error signal may lead to a refusing action or a mis-operation of relevant equipment, which will cause trouble to the mitigation of the earthquake accident of the nuclear power plant. Relay chatter could also actuate the lockout devices or seal-in devices associated with the supply breaker for essential buses, thus operator manual recovery is needed to re-close these breakers. Considering the accessibility of these equipment locations after an earthquake, it is difficult for operators to manually restore them if the number of these equipment is large. So relay chatter may cause an adverse impact on the accident mitigation functions by rendering selected accident mitigation equipment unavailable. This paper briefly introduces the method and procedure of Relay Chatter Analysis for the nuclear power plant PSA or SMA, then, illustrate it with an example of 380 VAC contactor control circuit. Finally, the modeling and quantification process for relay chatter analysis results in seismic PSA is briefly introduced, and a reference for solving similar problems is given.
An Evaluation of Critical Heat Flux Prediction in Single-Side Heated Narrow Rectangular Channel Under Upflow Condition
Authors: Yan Meiyue , Ma Zaiyong , Pan Liangming , He Qingche , Xv Wangtao , Li Xiang 
Published: November 2022

Abstract
Safety is the biggest concern of nuclear power plant and research reactors.As one of the three primary thermal design criteria for rectors, the critical heat flux (CHF) guarantee the reactor’s safety and economics. The bubble dynamics in narrow rectangular channel is didderent from that in conventional channels. Therefore the CHF characteristics obtained in the conventional channel may not be suitable in the narrow rectangular channel. To analyze CHF characteristics in the narrow rectangular channel, a visual experiment study on CHF was carried out in one side heated narrow rectangular channel under upflow condition condition. The experiments were performed at pressures ranging from 1 to 4 MPa, with inlet subcooling ranging from 65 to 120 K and mass flux ranging from 350–200 kg/(m2s). The deionized water was used as the working medium. The relevant thermal-hydraulic parameters and visualization results were collected synchronously. Then compare the CHF experimental values with the predicted values of the W-3 correlation,Mishima correlation, and Chang H correlation. The result shows the predicted values of W-3 correlation and Mishima correlation are always larger than the experimental values. Among these corrrelations, we found that the errors of Chang H correlation is within 30% in 2mm gap size.
Experimental Research of Bundles Radial Power Distribution Influencing Factor on Fuel Assembly Mixing Characteristics
Authors: Cheng Cheng , Donghua Lu , Qianhua Su , Chengyue Wang , Yanyan Xi 
Published: November 2022

Abstract
It can be seen that the studies on fuel assembly mixing performance in the reports are all focused on the influence factors of thermal parameters, and few on the structural parameters. In order to systematically investigate the influence of radial power distribution of rod bundles on the mixing characteristics of fuel assembly, the experimental study was carried out by a 14 ft 5 × 5 full-length rod bundle fuel assembly with both left-right symmetry and center-symmetric arrangement of cold and hot rod bundles. CFD numerical simulation method was used to calculate the same condition. The results show that the radial power ratio of the rod bundles has little effect on the mixing coefficient kT of the fuel assembly when the rod bundle is centrally symmetrical. When the rod bundle is left-right symmetrical, the temperature matrix of the sub-channel section presents an angular symmetry phenomenon near 45°. The experimental results and CFD simulation both show that the configuration of the mixing vanes forms a macroscopic diagonal circulation of the fluid in the sub-channel. FLLICA-IIIF subchannel software could not accurately simulate the diagonal circulation phenomenon of the fluid in the subchannel.
U Influence of Non-Condensable Gas in Mobile Heat Pipe Small Reactor on Heat Pipe Heat Transfer Performance
Authors: Tianyu Gao , Tao Zhou , Wenbin Liu , Jianyu Tang , Huaichang Lu 
Published: November 2022

Abstract
Heat pipe cooled reactors have a promising future in the next generation of reactor design. The heat pipe, which is the core device for heat transfer from the core to the outside, determines the safety of the heat pipe reactor and the efficiency of power generation. In the high-temperature, high-pressure, high-irradiation environment of a nuclear reactor, the non-condensable gas produced by the lithium heat pipe - helium - affects the heat transfer efficiency of the heat pipe. Therefore, by modeling the heat transfer process of the heat pipe, the effect of the content of the non-condensable gas on the heat transfer performance of the heat pipe is calculated. It can be concluded that: at a certain input power, the temperature of both the hot end and the cold end of the heat pipe increases with the increase of the content of non-condensable gas; the distribution of non-condensable gas in the heat pipe is irregular, more at the hot end and less at the cold end; at a certain input power, the equivalent thermal resistance of the heat pipe increases with the increase of the content of non-condensable gas, and the higher the content of non-condensable gas, the greater the effect on the heat transfer effect of the heat pipe; in a certain range The effect of power on the equivalent thermal resistance of heat pipe is almost none.
Application of Recurrent Neural Network in Improving Software Reliability Model of Reactor Protection System
Authors: Li Duo , Zhang Wenji , Guo Chao , Zhou Shuqiao 
Published: November 2022

Abstract
The safety-critical software of Reactor Protection System (RPS) plays a significant role for the safe operation of the Nuclear Power Plant (NPP). The reliability analysis of safety-critical software is also expected by the nuclear regulation agencies and the software development groups for test evaluation and optimization.
Software Reliability Growth Model (SRGM) is a commonly used reliability model. SRGMs can be divided into Markov model, non-homogeneous Poisson process (NHPP) model and so on according to the stochastic process method of its application. The NHPP model occupies a significant proportion in the software reliability models, which considers the failure rate as a function of time and, in the ideal case, the failure rate decreases with the removal of faults; after the software is released, the failure rate no longer changes unless the software is debugged again.
SRGM researches are based on the recorded software error data which are a set of time series data and collected during software development. There are lots of extensive and in-depth researches to improve the software reliability growth model, such as considering more hypothesis in software development, analyzing software error severity, and so on. The rapidly developed neural network technology provides a new kind of research tool to analyze these recorded data. This paper uses software error data to train a Recurrent Neural Network (RNN), the trained RNN can give more correct software reliability growth prediction and improve SRGM results.
Scaling of Thermal and Flow Behavior in Dry Storage Casks for Nuclear Spent Fuel
Authors: Wen Qinglong , Yang Ting , Xu Shijia , Wang Zhengquan , Yuan Bo 
Published: November 2022

Abstract
In China, the demands for nuclear power have increased with increasing energy consumption. Thus, requirements in the treatment of spent fuel have also increased. As spent fuel is radioactive, cooling is required to prevent the spent fuel from reaching high temperatures. From the perspective of long-term development, dry storage casks have become a priority intermediate storage method in China. Before this storage method is put into application, the nuclear regulatory authority requires a thorough safety evaluation by experiments or analyses to ensure that it maintains its thermal safety function under normal, abnormal, and accident conditions. It is pointed out that experiments with the full-scale model are expensive and are typically very large. However, the safety of such structures may be examined through the use of a dimensionally scaled-down model. To ensure an accurate analysis of the scaled-down structures, the scaling of thermal-hydraulic behaviors from the prototype dry storage casks for spent fuel is required.
Firstly, a general scaling method of a conventional testing system is introduced in this paper. Secondly, the principle used to perform the scaling of a dry storage cask simulator is carefully explained. Then a Phenomena Identification and Ranking Table (PIRT) of thermal and flow behaviors involved in a dry storage cask are presented. After that, dimensionless criteria which are used to characterize thermal-hydraulic behaviors in the dry storage cask are derived. Finally, the distortion is numerically analyzed.
Rejection of Synchronous Vibration Based on Generalized Adaptive Period Iteration Algorithm for Active Magnetic Bearing
Authors: Bian Xiaoyu , Shi Zhengang , Mo Ni , Zheng Yangbo , Shi Lei 
Published: November 2022

Abstract
High temperature gas-cooled reactor (HTGR) is the fourth-generation advanced nuclear reactor developed by Institute of Nuclear and New Energy Technology, Tsinghua University, which has advantages of inherent safety and modular design. Since the rotor is suspended by electromagnetic forces, active magnetic bearings (AMBs) possess the advantages of on friction and no need of lubrication. Hence the AMBs are widely applied in HTGR, which can ensure the pure helium environment for rotational machineries. However, in general, the imbalance of the rotor will result in the vibrations of the rotor and the shell of the machine during operation, which seriously affects the stability and security of the AMB system. This paper proposes a generalized adaptive period iteration (GAPI) algorithm. This method can estimate the rotating speed based on the displacement vibration signals and realize the rejection of the synchronous vibrations, which keeps the rotor rotating around its geometric axis. Unlike conventional unbalance control methods, the proposed algorithm does not require the rotating speed sensor, which simplifies the structure of the AMB system. Moreover, this method is independent of mathematical model of the system and can adoptively adjust the amplitude and phase of the feed-forward compensation signals to suppress the synchronous vibrations. Simulations and experiments show that this algorithm can quickly and accurately estimate the rotating speed, and has good stability and robustness.
Seismic Analysis of Fuel Transfer Tube Gate Valve
Authors: Ziwei Su , Xin Zhang 
Published: November 2022

Abstract
The Gate valve is an important functional part of the Fuel Transfer system, in order to achieve complete independence of Fuel Transfer unit, the scientific research and manufacture of the gate valve is imperative. According to the objectives, the overall structure design, component function design and material selection are completed, and then manufacture scientific research prototype, carry out a series of performance tests such as strength test, sealing test and whole machine life test, etc. Among them, it is an important step to verify the integrity of the Gate valve under seismic conditions, so in accordance with the design structure, establish the 3D geometric model, divide mesh, creat the finite element model, and apply the boundary conditions, the numerical simulation for Gate valve was presented by using ANSYS software.
The results show that under a combination of loads of various working conditions, including SL-1 abnormal condition or SL-2 accident condition, the stress of each component of Gate valve body, Support, Operating mechanism and Driving rod is less than the specified limit and meets the requirements of the judgement criterion; so the equipment achieve complete independence of Fuel Transfer unit, and be with reliable and stable performance, which can effectively guarantee the safe operation of nuclear power plant.
Design of Loading Elevator Control System Based on Servo Positioning Control
Authors: Zhao Lihui , Dong Shengguo , Zhu Hao 
Published: November 2022

Abstract
Loading elevator is one of the important equipment for refueling system. The refueling process of the sodium-cooled fast reactor is sealed and the equipment for refueling process is complicated and in large quantity. It is necessary to research and develop a set of high-accuracy and high-reliability control system to ensure loading elevator for the refueling process can operate safely and reliably. This article illustrates and describes loading elevator’s control system design detail. For meet the requirement that the positioning deviation is limited to ±5mm, this control system hardware for loading elevator adopts the advanced digital AC server to position the equipment. Compared with the traditional PLC control method, it has the advantages of high positioning accuracy, simple commissioning process and stable operation. Meanwhile, the multi-fault detection logic is used for fault detection. Though acquire real-time data, the fault logic transfers protection signals and interlocking signals to DCS. This fault logic can ensure that when the equipment operates abnormally, it can detect the fault in real time and stop the motor to ensure the safety of the equipment and components. The control system design plan has been applied in the loading elevator test bench safely and reliably for many times, which proves that it can realize the function of equipment monitoring and control, and also the positioning accuracy can meet all requirements.
Application of Qualitative and Quantitative Configuration Risk Assessment in AP1000 Unit Outage Scheduling Optimization
Authors: Wang Zhenqi , Huang Weibing , Zhou Jian , Lin Huan , Tao Yinyong , Zhang Jun 
Published: November 2022

Abstract
Configuration risk management refers to the use of probabilistic safety assessment technology to calculate risk indicators and perform risk management based on the actual operation configuration of nuclear power plant. Since its development in 1990, this method has been widely used in nuclear power plants. It can help power plant personnel optimize maintenance plan scheduling, control the nuclear risk level of units and improve the safety of power plants. During the shutdown and outage of the nuclear power plant, a lot of work has to be performed, resulting in the centralized shutdown of system equipment, and the nuclear safety during this period has attracted the attention of the power plant. Therefore, in order to ensure that the nuclear risk of the unit shutdown and outage can be controlled, Sanmen Nuclear Power, as the world’s first AP1000 unit, actively performs configuration risk management and control during the shutdown, It is the first time in China to optimize the outage plan scheduling through the combination of qualitative and quantitative configuration risk assessment method, so as to comprehensively investigate and eliminate the potential risk sources. This paper will mainly introduce the development background of configuration risk management, and take Sanmen Nuclear Power AP1000 unit as the object to introduce the application and achievements of qualitative and quantitative configuration risk assessment in outage scheduling optimization. In addition, it will also make a comparative analysis of qualitative and quantitative configuration risk assessment techniques.
Design of Radiation Damage Monitoring Scheme for Materials in China Experimental Fast Reactor
Authors: Ziao Hu , Xiao-xian Chen , Xiao-liang Chen 
Published: November 2022

Abstract
In order to monitor the radiation effects of advanced fuel element cladding materials, it is necessary to install irradiation supervision tanks in the reactor, which contain kinds of detector foils. And then the detector foils are used to determine the neutron energy spectrum and neutron flux, so as to evaluate the radiation damage of the material. Taking China Experimental Fast Reactor (CEFR) as the research object, seven neutron detector foils were selected and five supervision schemes were designed in 316 stainless steel. The size of the neutron detector, the irradiation time and the corresponding relationship with the radiation damage are calculated by the Monte Carlo code and SPECTER program, and finally the sound monitoring scheme is given. It has guiding significance for irradiation experiments of structural materials, and provides a basis for monitoring the irradiation damage of large commercial fast reactor structures.
Multiplication Factor Prediction Feasibility Study for the BEAVRS Core Based on the LSTM Algorithm
Authors: Chang-an Ren , Ji-chong Lei , Xiao-hua Yang 
Published: November 2022

Abstract
This paper focuses on exploring the feasibility of the LSTM (Long Short-Term Memory) algorithm in deep learning for effective multiplication factor keff prediction at the core level, modeled by BEAVRS (Benchmark for Evaluation And Validation of Reactor Simulations) core first cycle loading with keff of operating at full power for 0–300 days was used as the study subject. The first 65% of the dataset is the training and validation set, and the last 35% of the dataset is the prediction target. The training and alignment results of the physical parameters of the components were obtained using the DRAGON4.1 and DONJON4.1 codes, and the LSTM algorithm in deep learning was applied. By adjusting the number of LSTM cells, L2 regularization parameters, optimizer type, and other parameter coefficients in the algorithm, the results showed that the absolute error of the predicted core effective multiplication factor keff could be made within 2 pcm by adjusting the appropriate parameters, which validated the successful application of machine learning to transport equations.