bluuebunny/crossref_metadata_2025_split huggingface dataset:
A Modified Simulation Model of Blocked Regions After Rod Cladding Deformation During LOCA Transient
Authors: Cao Zhiwei , Lin Zhikang , Xu Miaomiao , He Jimin , Song Jianyang 
Published: November 2022

Abstract
During a loss-of-coolant-accident (LOCA) transient, the ballooning and burst of the cladding can lead to blockage of flow path and worsen core coolability. Studies have shown that the core coolability is not changed much up to about 60% blockage ratio, while can be threatened by a severe blockage with a high blockage ratio and long blockage length. A blockage simulation model based on the non-coplanar assumption is widely used in LOCA transient analysis. However, regarding the coplanar deformation observed in experiments, validity of this assumption is questionable. Furthermore, it is found that this model cannot predict reasonable results for severe blockage conditions.
In this paper, the blockage model is modified for the thermal hydraulic simulation of LOCA transients with the occurrence of severe blockages. Pre-test simulations of a test facility studying the influence of blockage on core coolability have been conducted with the modified model. The results show that, the modified model shows better agreement with the observations in the experiments.
Study on Reactor Vessel Water Level Measurement Method for Emergency Decision-Making of Nuclear Power Plant
Authors: Shi Yanming , Wang Zhenying 
Published: November 2022

Abstract
After a nuclear accident, the nuclear power plant will monitor and analyze the unit status, environmental characteristics and accident process, determine the emergency state classification under the nuclear accident conditions, and respond to the emergency state hierarchically, to reduce the impact of the accident and ensure the safety of personnel. For determining emergency state classification of nuclear power plant cold shutdown and refueling shutdown operation mode, reactor vessel water level is an important basis for judging. The Three Mile Island nuclear accident made the industry realize the necessity of monitoring the primary water load, especially the water level in the Reactor vessel. Reactor vessel water level provides an essential basis for monitoring the core cooling state after an accident, to ensure that the core cooling state can be diagnosed correctly in time, accurately and conveniently, thus providing a criterion for determining the emergency state level, and then selecting the appropriate accident operation strategy.
Based on the demand analysis of emergency state classification based on reactor vessel water level, combined with the current situation of reactor vessel measurement in Reactor vessel and the setting of peripheral dose rate monitoring channels, this paper supplements the water level measurement method under potential water loss conditions, realizes the full-range measurement of Validity water level in Reactor vessel, evaluates Validity water level Validity in Validity, and then efficiently determines the emergency state classification, providing Validity basis for emergency response.
A Study on Health Assessment Technology for Sensor of Nuclear Power Plant Based on Dissimilarity of Residual Distribution
Authors: Li Wei , Zhang Yiwang , Yuan Yidan 
Published: November 2022

Abstract
In this paper, we found that the distribution of residuals (i.e., the difference between the estimated value of state estimation model and the sensor reading) of a sensor in abnormal state would change significantly comparing with the sensor in normal state. So, we proposed a health assessment technology based on the dissimilarity of residual distributions. The distance between these two distributions can be used as the health indicator of the sensor. The greater the distance, which means the less similar the two distributions are, the worse the health status of the instrument. The calculation results showed that the dissimilarity of residual distributions had certain universality as a health indicator of sensor, and it could accurately reflect the health state of a sensor. This technology can judge the degradation of the sensor in advance through the real-time online health assessment. By using this technology, the safety of nuclear power plant can be improved and the operation & maintenance cost of nuclear power plant can be reduced. In the later stage, the combined distribution of signal residuals from multiple sensors can also be used to evaluate the health of a target system.
Optimization of the Tritium Production and Discharge in HPR1000
Authors: Fu Pengtao , Dai Mingliang 
Published: November 2022

Abstract
Tritium is one of the most important radioisotopes discharged into the environment in the reactor and contributes to 99% of the total radioactivity of all radionuclides in the liquid discharges during the normal operation of pressurized water reactors. The discharged tritium can be taken into the human body via drinking water. Therefore the minimization of production and discharge of tritium has been highly focused on in the industry and the public across the world.
In the paper, all the sources of tritium generation in HPR1000 have been introduced. Activation of Boron-10 in the primary loops and Beryllium from secondary neutron sources have been recognized as the two main sources of tritium in the primary coolant in HPR1000. The annual tritium production and discharge in HPR1000 have been estimated to be about 4.2E+01 TBq/yr using CGN OPEX data with appropriate corrections to reflect tritium-related differences between HPR1000 and the CGN units. To minimize tritium discharges as far as possible, the feasibility of eliminating the secondary neutron source assemblies from the third cycle in the HPR1000 has been verified and the annual tritium can decrease by up to 48% correspondingly. The method used in the study has been certified by UK Generic Design Assessment and EUR compliance assessment. This approach can also be applied to the minimization of tritium discharges in operating PWR units and in the new PWR units.
Study on Time-Dependent Co-58 and Co-60 Activities in the Primary Coolant of CPR1000 PWRs
Authors: Zhang Xiaoqian , Fu Pengtao 
Published: November 2022

Abstract
Activated corrosion products are the dominate radiation source during PWR maintenance. Co-58 and Co-60 are two important radionuclides due to the high energy emitting gamma rays and high radio activities in the primary circuit. They have contributed to more than 90% occupational radiation exposure during maintenance according to operating experiences from PWRs. Therefore, it is essential to research the behaviour of the two radionuclides.
The Co-58 and Co-60 coolant activity for both short-term tendency and long-term tendency are analyzed in this paper. Operating Experience (OPEX) from CPR1000 PWRs show that the Co-58 and Co-60 coolant activity in a fuel cycle usually result in shape of a bowl and the highest activities are at the beginning of the cycle which is consistent with the shape of the core boiling rate. Namely, fuel crud behavior plays a leading role in the formation of activated corrosion products in the primary circuit. For Co-58 and Co-60 coolant activity in long-term period, it is influenced mainly by the corrosion/release rate and the radionuclides’ half-lives. OPEX combined with simulation by CAMPSIS code show that Co-58 coolant activity reach to maximum value at second to third cycle while Co-60 coolant activity reach to maximum value at fifth to ninth cycle. Research on time-dependent Co-58 and Co-60 coolant activities give a better understanding of the radionuclide behaviour as well as provide a basis for developing corrosion product simulating model.
Concerns Regarding Equivalent Substitution of Entity Structures by Porous Media Models
Authors: Bai Yufei , Zhang Wei , Yao Yangui , Yao Weida 
Published: November 2022

Abstract
Reactor vessels and steam generators are characterized by complex structures. Because of the limitation of computing resources, simplifications are necessary investigating the internal flow and heat transfer characteristics. The commonly adopted method is substituting reactor core, heat transfer tubes and support plates by porous media models. In order to investigate the differences between porous media models and entity structures, computational fluid dynamics (CFD) is applied to perform numerical simulation of flow through an orifice plate. The pressure drop, velocity and flow distribution are compared under various conditions. The porous media model is further set downstream the orifice plate to study the influence of distance on the interaction of two throttle elements. The results show that the flow field in the vicinity of the throttle element varies significantly. The local loss is observed when the velocity distribution uniformity changes. However, evenly distributed resistance is generated in porous media models. Therefore, the internal flow distribution is uniform and the pressure drop is only generated inside. When the two throttle elements are close, the pressure drop of each throttle element and the total pressure drop are influenced due to the fact that the downstream flow of the upstream throttle element is not fully developed. The influence of the interaction decreases with the distance. The two throttle elements could be decoupled when the flow is fully developed. It should be properly considered to make sure the resistance calculation is valid and conservative.
Stable Feature Reweighting Transformer for Out-Of-Distribution Bearing Fault Diagnosis
Authors: Chen Hao , Nie Jie , Yu Shusong , Wang Chenglong , Yin Peizhe , Yang Kun 
Published: November 2022

Abstract
Rolling bearings play an important role in the aerospace industry, manufacturing, and nuclear engineering. To ensure the reliable and stable operation of various mechanical equipment, the research on bearing fault diagnosis is very practical and critical. With the rapid development of smart manufacturing and industrial big data, deep learning has become an effective solution for emerging fault identification. Due to the different distributions of training samples and test samples for bearing faults, researchers have introduced many transfer learning methods to solve this problem. The traditional methods assume that the features of the target domain are known, or that the target domain has sufficient samples (similar to the number of training samples). However, these conditions are often difficult to meet in reality. In this paper, we proposed a Stable Feature Reweighting Transformer (SFRT) to deeply mine correlated features in fault signals and remove dependencies between correlated and irrelevant features. This helps the model mine the correspondence between distinguishing features and fault labels. Therefore, our method can complete the fault diagnosis task under variable working conditions with high quality when the target domain is unknown. Extensive experiments on CWRU and PU datasets demonstrate the effectiveness of our method on multiple distributional generalization tasks. Compared to state-of-the-art methods, our model produces the best fault identification accuracy.
Analysis and Simulation in Rotating Rectifier in Brushless Excitation of Nuclear Power Plant
Authors: Xingzhen Zhang , Hui Sun , Yanjun Zhao , Xianzuo Shi 
Published: November 2022

Abstract
In recent years, a number of projects under construction in nuclear power plants have successively experienced excitator failures at key nodes of grid connection and commercial transportation, resulting in downtime, which has had a great impact on the progress and operation performance of the unit. The rotary rectifier bridge is an important part of the generator excitation system and is the most important link in the final realization of the rotor excitation. As a rotary rectifier equipment, it is difficult to achieve online real-time monitoring, the current large-scale nuclear power unit rotary rectifier bridge online real-time monitoring results are not ideal, there is on-site feedback false alarm, the need for high maintenance costs. With the gradual increase of nuclear power units, it is particularly important to develop a real-time online monitoring device for brushless rotary rectifier bridges and to achieve early warning. The equipment can win valuable time for the operation and maintenance of the unit, even if it needs to be protected indirectly by other protections in the event of failure, it can also provide direct data for post-accident problem analysis, locate the fault point, reduce the problem analysis time, improve the maintenance efficiency, and have great economic benefits.
Research on Automatic Arrangement Method of Process Piping in Nuclear Power Plant
Authors: Su Jincheng , Xue Jing , Jia Xiaopan , Gu Zhenjie 
Published: November 2022

Abstract
According to the process and content of process piping layout design, based on the A-star algorithm, this paper introduces the concept of automatic arrangement, and proposes an automatic arrangement design method for nuclear power plant process piping. The method takes into account several factors, such as the nuclear power plant plant scheme, the connection relationship and location of process equipment, the constraints involved in the design of process piping layout, and the performance of the process piping layout scheme. Based on this method, the programming simulation of the automatic arrangement of the nuclear power plant process piping is carried out, the automatic arrangement design of the nuclear power plant process piping is realized, and the applicability of the method in the automatic arrangement of the process piping of the nuclear power plant workshop is verified.
Optimization Configuration and Design Considerations of Spent Fuel Storage and Transport Casks for Nuclear Power Plants in China
Authors: Liu Fengwu , Li Shiyuan , Wang Lijun , Jing Shunping , Liu Chao , Li Xiaoqing 
Published: November 2022

Abstract
With the rapid development of nuclear power in China, the amount of spent fuel keeps increasing, which makes the demand for safe and suitable casks to store and transport spent fuel more prominent. Due to China’s national policy of “closed nuclear fuel cycle” and the uncertainties in the construction of reprocessing facilities, the configuration of spent fuel storage and transport cask has been directly related to the safety of nuclear power plants operation. Based on the demand of spent fuel storage and transport in China, this paper analyzes the requirement of different modes of storage and transportation and different loading contents for casks, such as 12ft / 14ft spent fuel loading, compatibility with short-term storage and transportation, and off-site transportation of dry storage canisters, etc. In order to ensure safety and control cost, this paper also analyzed the applicable standards, transport restrictions and optimization design objective for the optimal design of spent fuel storage and transport cask, and puts forward the optimal configuration and design method of the storage and transport cask based on the analysis of key concerns, which could provide reference for the configuration and design of spent fuel storage cask.
Application of Industrial Digital Twins in Nuclear Power Plant
Authors: Xie Jing , Fu Yueming , Zhang Xun , Zhang Hengdi , Hou Jianfei , He Xiangjie , Yu Fei , Lin Run 
Published: November 2022

Abstract
Industrial digital twins have become popular in China and around the world in various fields in the past a few years. The application of industrial digital twins in nuclear power plants is to be researched in this paper.
This article mainly summarizes the concept of industrial digital twin and its application status; sets up an application system of industrial digital twin in nuclear power plants, including the overall framework and application mode; introduces some typical application of digital twins in nuclear power plants; explores the expectation, challenges of the application of digital twins in the development of future intelligent nuclear power plant.
Study on the Method for the Safety-Related Instrument Calibration Periodic Test Interval Extension of Nuclear Power Unit With Digital I&C System Based on Setpoint and Uncertainty Analysis
Authors: Zhang Zhao , Zhang Yong , Wang Zi-Chun , Tang Bo-Wen , Zhou Yong-Hui , Su Min-Hua , Luo Wen-Bo 
Published: November 2022

Abstract
Extending the intervals of safety-related Instrument Calibration Periodic Test can shorten the refueling outage period and improve unit economy. Based on the general requirements for extending the Safety-related Instrument Calibration Periodic Test Intervals and in consideration of the characteristics of CPR1000 Nuclear Power Unit with Digital Control System, an evaluation method is proposed and investigated, which consist of instrument drift analysis, protection channel analysis, channel uncertainty calculation, setpoint analysis and relevant channel tests evaluation. The research on the extension of pressurizer pressure transmitter calibration test interval is carried out in detail. The results show that the drift resulting from the interval extension will have no effect on the protection channel setpoint, the acceptance criteria for channel consistency test should be reduced and the acceptance criteria for channel function test should remain unchanged. The research of this paper has a reference value for unit economy improvement and outage optimization.
Research on Periodic Test Optimization of Digital Instrument and Control System in Nuclear Power Plant
Authors: Gao Qiang , Yang Xue , Li Qiuqi , Xiong Wenbin 
Published: November 2022

Abstract
The digital instrument and control system has been widely used in nuclear power plants as the power plant control system. The periodic testing of digital instrument and control system is regularly used to ensure that the digital instrument and control system can correctly perform its functions according to the design requirements and keep in the available state required by the operating technical specifications, so as to maintain the reliability of the system. During the actual operation of the nuclear power plant recently, it is found that some periodic tests continue the test content or method of the traditional analog control system, and some periodic test contents are not suitable for the actual site or the test frequency can be optimized. In this paper, the periodic test is optimized from the following aspects: the periodic test frequency of the digital instrument and control system, the coverage of the test content, the experience feedback of the nuclear power plant, the introduction of risks to the periodic test, the reliability of the instrument and control system, and the risk guidance technology. And form optimization suggestions to ensure that the digital instrument and control system can give full play to its advanced nature under the premise of safety and availability.
High Temperature Gas Cooled Reactor Steam Generator Research on Stress-Rupture Test of High Temperature Materials and Welded Joints
Authors: Bao Shimei 
Published: November 2022

Abstract
The outlet design temperature of the steam generator of high temperature gas-cooled reactor is 570°C, and the design temperature of the high temperature components have reached or exceeded the creep temperature of the material. In the creep temperature range, the allowable stress of the material mainly depends on its high temperature stress-rupture strength. Therefore, it is necessary to High-temperature stress-rupture test studies are carried out on materials and welded joints to verify whether their stress-rupture strength can meet the design requirements. High-temperature stress-rupture tests were carried out on 9 kinds of high-temperature materials and 5 kinds of welded joints by designing the non-standard specimen for heat exchange tubes, reasonably setting test stress points, parallel samples and sampling positions.14 double logarithmic curves of stress-rupture time were drawn, and the long-term high-temperature stress-rupture strength of high-temperature materials and welded joints was extrapolated, Providing important data support for the safety evaluation and life evaluation to the steam generator of high temperature gas-cooled reactor.
Experimental Study on Droplets Entrainment and Film Breakup Characteristics Under in Rotating Churn Flow
Authors: Kang Ruiqi , Xiong Zhenqin , Li Zhen , Ouyang Shuo , Liu Jing 
Published: November 2022

Abstract
The separation performance of swirl-vane separators, including pressure drop and separation efficiency, is closely related to the two-phase flow pattern and the liquid film breakup behavior. In the swirling-swirl-vane separators, the droplets collision and film breakup happen. Influenced by the spiral blade structure and the central hub, the swirl-vane outlet region is prone to the formation of churn flow. Due to the complex and variant thickness of the liquid film in churn flow, it is extremely difficult to visually observe the generation and development process of the droplet entrainment from the side or the top under normal working conditions. Data on the phenomenon of droplet entrainment under the rotating flow field are rare, resulting in the entrainment mechanism of droplets under the rotating flow field is unclear. In this paper, investigations on the entrainment of droplets downstream of the swirl-vane has been carried out for a small liquid flow rate. A rotary flow field visualization experimental device has been developed. The churn flow is formed under the condition of near-zero liquid flow. The motion characteristics of the droplets in the churn flow are extracted by observing the movement of a small number of droplets in a rotating flow field. The rotational churn flow and droplet entrainment process are obtained through visual photography. The liquid film ligament and bag breakup characteristics under the rotating flow field are obtained, as well as the particle size characteristics of the entrainment droplets.
Performance Evaluation of Boron Doped NMCP Used in Neutron Radiography
Authors: Tan Jinhao , Zhou Jianrong , Song Yushou , Yang Wenqin , Zhou Jianjin , Xia Yuanguang , Liu Hui , Liu Shulin , Yan Baojun , Wang Songlin , Sun Zhijia , Chen Yuanbo 
Published: November 2022

Abstract
Neutron radiography plays an important role in the development of safer and more efficient nuclear power plants by providing valuable information for the irradiated nuclear fuel, nuclear waste containers and other critical components. Neutron sensitive microchannel plate (nMCP) doped with boron and gadolinium is a popular neutron imaging detector. It has been successfully employed in neutron resonances imaging of irradiated nuclear fuel, providing the possibility in mapping the elemental composition of reactor components. However, the exiting of gadolinium in nMCP degrades the spatial resolution and neutron-gamma discrimination capability, because of the long range and low energy deposition of the electrons generated by gadolinium. nMCP doped with boron has advantages on spatial resolution and neutron-gamma discrimination capability. In this paper, performances of boron doped nMCP were studied. According to the simulation, the theoretical detection efficiency of nMCP doped with 20mol% 10B2O3 could reach 34%, and the spatial resolution can be close to the pore size at the same time. Moreover, it is expected to achieve good neutron-gamma discrimination ability. The detection efficiency for nMCP which doped with 10mol% natB2O3 was experimented detailly at CSNS. Results show that the detection efficiency is 4.55% for 2.9 Å neutron and 7.26% for 4.8 Å neutron, which agrees well with simulation results. These results indicate that the excellent performance of boron doped nMCP is possible to be achieved. In the future, nMCP doped with 20mol% 10B2O3 will be fabricated and it will be a promising scheme for the application of neutron radiography in nuclear reactors.