bluuebunny/crossref_metadata_2025 huggingface dataset:
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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Non-Destructive Examination of PWR Steam Generator Tube-Tube Plate Welds After Repairing Melting Damage
Authors: Zhang Yan , Zhang JianLei , Cheng Zhonghe
Published: November 2022

The tube-tube plate weld is the boundary weld that separates the primary and secondary circuits of the Pressurized Water Reactor (PWR) Steam Generator and is the weakest link of the Steam Generator. If the weld leaks, the media that may be contaminated by radiation in the primary circuit may flow into the secondary circuit, resulting in a nuclear leakage event. Therefore, the welding quality of the weld is directly related to the safety of the nuclear power site and requires special attention during the manufacturing process. The tube-tube plate weld of the PWR Steam Generator generally uses non-filled wire self-fusion welding technology. In addition to common welding defects such as pores, incomplete penetration, and cracks during the welding process, melting damage may sometimes occur due to improper operation. Since the melt loss repair involves both the tube plate surfacing layer and the tube-to-tube plate sealing weld, the repair process is much more complicated. There is little experience in the successful implementation of non-destructive inspection after melting damage repair. How to remove the defects? Which non-destructive examination methods should be selected in the repair process to ensure the defect removal effect? How to weld repair after the defect removal? Which nondestructive examination techniques should be selected during the repair process to ensure the repair effect? How to conduct final non-destructive testing after repair to confirm Rework quality? These are urgent issues. In this paper, by using tube-tube plate weld samples to carry out melting damage repair simulation experiments. After repairing, the penetration examination and the radiographic examination have been successfully implemented, and good results have been achieved.
Research and Development of Three-Dimensional Digital Operation Management Platform for NPPs
Authors: Wang Hao , Tian Li-Cheng , Lin Jia , Huang Hui-Zhen
Published: November 2022

With rapid development of informatization and digitalization technologies, digital transformation is pushing ahead in industry around the world, especially in nuclear power operation and maintenance. In the full life-cycle digital transformation plan of nuclear power enterprises, the design institute has accumulated a large number of digital design results, such as the three-dimensional model data of nuclear power plants. These 3D models are handed over to the operation organizations to build digital, visual and intelligent operation and maintenance solutions. Without digital models, field staffs have to check the paper procedures and technical data before or even during the on-site inspection and maintenance preparation, however, this is adverse for field work. The application of 3D digital auxiliary tools and platforms can effectively improve management efficiency and reduce potential risks in nuclear power operation and maintenance. This paper proposed a digitalization and visualization method to improve the digital management level, and developed a 3D visual operation management platform for Daya Bay nuclear power plant. With the platform, nuclear power operation organizations could handle the daily maintenance work in an efficient way, so that to improve the operation management level.
Research on Key Problems of Digital Instrument and Control System Reliability Analysis for Probabilistic Safety Assessment
Authors: Lin Shuhan
Published: November 2022

Under accident conditions, the support of digital instrument control system especially the normal operation of protection and safety monitoring system has an important impact on the safety of nuclear power plants. As the basic core components of digital I&C system, the card modules integrate and cooperate to perform nuclear power plant operation, data display, control, protection and safety monitoring functions. The existing card reliability analysis methods appeared to be too conservative, and lack enough operation experience and field test as sample data. In this paper, analysis tools such as RiskSpectrum PSA professional software are used, through the methods of static fault tree model refinement, card reliability test, Bayesian/frequentist statistical analysis, a quantification method of card reliability parameter are put forward, which is more suitable for PSA. The PSA model of digital I&C PMS is refined and optimized according to the results, truly reflects the reliability level of the digital I&C system. While verifying the advancement of digital I&C system, this work provides technical basis for design and operation revision or optimization decision, thus improving the safety and economy of nuclear power plants.
Development of a 1d Code for Heat Transfer Characteristics of Helical-Coiled Once-Through Steam Generators
Authors: Liu Maolong , Xu Ziyi , Xiao Yao , Gu Hanyang
Published: November 2022

The helical-coiled once-through steam generator (H-OTSG) is a key equipment in small modular reactor (SMRs) to be used in heat exchanging between primary side and secondary side. In this study, a 1D Code HTAC (H-OTSG Thermo-hydraulics Analysis Code) is developed to simulate the heat transfer characteristic of H-OTSG and used to prepare the experimental study in future. The H-OTSG is composed by 74 helical tubes which is divided into 10 layers according to the coil diameters. Firstly, a 1D code of H-OTSG is developed with the tube-side thermal-hydraulic model proposed by the present authors and the shell-side model using the thermal-hydraulic model of straight-tube bundle in cross flow. Secondly, 10 inclined straight pipes with different coil curvature ratio are stablished in the 1D code to respectively simulate the helical tube bundle of 10 layers and the shell-side of H-OTSG is model with a vertical annulus. Furthermore, the sensitivity study on the node number is conducted and 99 nodes for each pipe and annulus are fine enough. Thirdly, the simulated result shows the 1D code can used to predict the steady-state heat transfer rate of the H-OTSG under various thermal power. Finally, the sensitivity analysis of the system parameters on the heat transfer rate of H-OTSG is conducted, and the result shows the decrease of tube-side pressure and the increase of shell-side inlet temperature can rise the average HTC of H-OTSG.
Studies on the Principles for Developing In-Service Inspection Acceptance Standards of Reactor Pressure Vessels
Authors: Pan Jun , Liu Chang , Su Xihui
Published: November 2022

ASME Code Section XI IWB-3500 gives the acceptance standards for in-service inspection of reactor pressure vessels, according to which the acceptable ratios of depth-to-thickness (a/t) varies along with depth-to-length ratios (a/l) for defects detected by nondestructive examination of reactor pressure vessels during the in-service stage. The acceptance standards are established based on fracture mechanics and five principal safety criteria as per EPRI NP-1406-SR, among which the safety margin between “allowable indication” in ASME Code Section XI and maximum postulated defects meeting ASME Code Section III fracture mechanics requirements is of most concern. The establishment of the acceptance standards can facilitate the development of new allowable indication standards to determine the component’s adequacy for continued service.
Quantitative Risk Assessment With CMMC Method on Abnormal Snowfall Incident for a Sodium-Cooled Fast Reactor
Authors: Nakashima Risako , Koike Akari , Sakai Takaaki , Doda Norihiro , Tanaka Masaaki
Published: November 2022

In general, the ultimate heat sink of a decay heat removal system is the atmosphere for accidental conditions of sodium-cooled fast reactor (SFR) designs in Japan. Therefore, risk assessment of external hazards from the atmosphere is important for SFR. However, along with global warming (GW), the number of significant meteorological disasters has increased. Therefore, it is necessary to focus on meteorological phenomena. Hazard curves were developed to address the possibility of abnormal snowfall. By using the hazard curve data, plant dynamics analyses by continuous Markov-chain Monte-Carlo (CMMC) method were conducted with the assumption of failure due to snowfall. The excess frequency of the coolant temperature was quantitatively examined for the design limitation temperature value. The abnormal snowfall height was evaluated using meteorological data. Assuming the possibility of snowfall increase occurring by 2050, the expected value to reproduce 10,000 years was evaluated, and the plant dynamics analyses were conducted using the snowfall height. In conclusion, abnormal snowfall was found to likely increase by GW. In addition, the CMMC method was implemented using the expected value to reproduce 10,000 years, and its effect on the probability of exceeding the temperature limit as the core damage factor was evaluated. As a result, the probability of exceeding this limit increased when GW was considered.
Analysis Method to Predict Grid-to-Rod Fretting Wear of Fuel Rod in Pressurized Water Reactors
Authors: Zhu Libing , Zhao Shufeng , Sun Jingya , Xu Shiyin , Gan Fujun , Li Sirui
Published: November 2022

Grid-to-rod fretting (GTRF) caused by flow induced vibration is one of the most important mechanisms of fuel rod failure in PWRs, which has a strong impact on the economy and safety of nuclear power plant. To avoid rod failure due to GTRF, several verification tests which are high costs and long term are always carried out for new design fuel assemblies. So as to deal with the disadvantages of tests and speed up progress in fuel design, a progressive analysis method to predict GTRF performance of fuel rod is developed, and systemic tests are carried out to verify the analysis method. The method consists of three main parts, including flow field analysis, structure vibration analysis and wear analysis. In the first part, the flow field excitation force is obtained by the large eddy simulation method, and this CFD method is verified by a particle image velocimetry (PIV) test with a 5 × 5 fuel bundle. In the second part, to get the vibration response of fuel rod, a nonlinear vibration model which can consider end-of-life gaps is established. Full scale fuel rod vibration in air and water conditions are tested to validate the predicted fuel rod response from analysis. In the third part, the fuel rod wear is estimated by tribological theory with fuel rod vibration response, the wear coefficient and the typical feature of wear scar. The GTRF method is finally verified by a long-term wear test of full-scale fuel assembly. It turns out that this method can effectively predict the wear characteristics of fuel rod cladding at the early stage of design. It provides a feasible solution for rapid design iteration of fuel assembly, and could reduce the development costs and risks.
The Test of Leakage Current of Terminal Stator Winding of the 4# Generator and Defects Treatment
Authors: Tian Kunpeng , Li Ping , Zhao Yu , Fei Xingwei
Published: November 2022

The insulation status of generator stator winding is related to the safe and stable operation of generator. When we repair the generator, we need different testing methods to verify the insulation of the different positions of the generator stator winding. In recent years, there is many short-circuit faults because of the insulation defect on the terminal of the stator winding. In generally, when generator need to be repaired, we do the AC/DC withstand voltage test. But when the results of AC/DC withstand voltage test are both OK, the insulation of the stator terminal winding is still not OK. that is the proof that the AC/DC withstand voltage test is not all-powerful. The test of leakage current of terminal stator winding is good effective method to checkout the the insulation of the stator terminal winding. We can find the insulation defect on the terminal of the stator winding with this test and avoid affecting the production even equipment damage.
Patent Perspective on the Development of Metal Nuclear Fuel Technology
Authors: Zhong Haoliang , Si Yu , Chen Xiaofei , Zhang Yading , Su Ran
Published: November 2022

Irradiated growth and irradiated swelling have been the key factors limiting the technological development of metal fuels. Compared with other types of nuclear fuels, metal fuels have the advantages of high thermal conductivity, high fission atomic density and easy processing, so the technological development of metal fuels has been of great interest to researchers, and there has been uninterrupted research and development over the years to overcome the technical problems of irradiation swelling of metal fuels. In this paper, we take a patent perspective and conduct a global patent search on uranium-zirconium alloys, uranium-molybdenum alloys, uranium-plutonium-zirconium alloys and metal fuels of other compositions in the technical fields of composition ratio, preparation process, preparation devices, fuel performance, fuel testing and back-end treatment of fuels, and analyse in depth the global development history of metal fuels, the patent situation of key applicants, the global sources of metal fuel technologies, the The analysis of the legal status, the analysis of key patent technologies, the technical means to solve the problems of irradiated growth and swelling, and the R&D direction of metal fuels at the present stage, and explore the future technological development path of metal fuels. Through in-depth analysis and research, this paper puts forward reasonable suggestions on how to break through the bottleneck of metal fuel technology and the follow-up development of metal fuel related technologies by designing and manufacturing fuel with special structure, improving the composition of metal fuel, modernizing and controlling processing methods, and improving processing technology.
Study on In-Containment Refueling Water Storage Tank Filter Upstream Effects Analysis Based on Three-Dimensional Design
Authors: Zhu Minghua
Published: November 2022

After the high-energy pipeline break accident in the containment, the pipeline insulation and other materials near the break will be damaged after impact, which will produce a large amount of debris. Some debris will be produced, a fraction of this dislodged insulation and other materials will be transported to the containment floor by the steam/water flows from the break and the containment spray. The in-containment refueling water storage tank is widely used in the third-generation nuclear power plant as the water source to support long-term recirculation for the functions of residual heat removal, emergency core cooling, and containment atmosphere cleanup. Typically, a filter is installed around inlet of pump to filter the debris and to minimize the amount of debris entering the downstream lines and the reactor core. The accumulated debris on the screen of filter may increase the differential pressure across the screen and thus decrease the net positive suction head margin available to pumps. The system may even disable all operational capability. In order to evaluating the performance of filters during and after accident, a reasonable conservative debris evaluation is required as design input. In order to accurately evaluate the impact of filter after accident, a reasonable and conservative debris source term evaluation is needed as the design input. The nuclear island design of the third-generation nuclear power first project fully adopts 3D modeling. This paper introduces the improvement of the upstream analysis method of IRWST filter based on three-dimensional design.
Research and Practice of Living PSA Maintenance Methods
Authors: Yang Zhi , Wang Chunhui , Li Gang , Wang Shihui , Pei Liang
Published: November 2022

In the technical policy for configuration risk management of nuclear power plants (Trial), it is clearly required to use the active probabilistic safety analysis model to carry out configuration risk management. Living PSA is not a new concept independent of PSA, but a higher requirement for PSA, to ensure that PSA truly reflects the design and operation characteristics of nuclear power plant, and truly reflects the risk of nuclear power plant. The development and maintenance of living PSA involves all aspects of nuclear power plant and requires close cooperation of multiple departments. Therefore, the construction of living PSA system requires not only reasonable maintenance methods, but also appropriate organization and quality assurance support.

Based on this point, this paper introduces the basic concept, content and development status at home and abroad of living PSA, studies the maintenance methods of living PSA, and puts forward the common problems and development direction in the maintenance process of living PSA in combination with the actual process and mode in the application process of CGN group, which has reference significance for the development of PSA model maintenance strategy, configuration risk management and maintenance rules.
Optimization of Transient Control Method for RCV System
Authors: Huang Yong Qiang , Chen Zhi Lin , Chen Xin , Gu Zheng , Chen Han Min , Zhang Run Ze
Published: November 2022

This paper provides an optimization scheme for the steam chamber of the pressurizer of the nuclear power unit. The scheme includes: if the operation information of the pressurizer meets the triggering conditions of the chamber, raise the charging flow and increase the discharge flow of the RCV system; Adjust the opening of the low-pressure control valve according to the measured pressure downstream of the letdown orifice; Close the connection control valve; If the discharge flow drops to the first preset flow, reduce the opening of the low-pressure control valve; Control the pressurizer to reduce the rising speed of charging flow; If the discharge flow drops to the second preset flow, stop the opening measurement adjustment of the low-pressure control valve and increase the opening measurement of the connecting control valve; If the letdown flow rises to the third preset flow, stop the opening adjustment of the connecting control valve, reduce the charging flow to the fourth preset flow, and control the pressurizer to perform cavity building operation. The method can effectively control the temperature rise of the fluid and the charging speed exceeding 107 °C. The service life of the nuclear power unit is improved, and the design of increasing the letdown flow to the third preset flow and reducing the charging flow to the fourth preset flow makes the letdown flow and charging flow meet the flow state of the pressurizer cavity, and ensures the stability of the pressurizer cavity establishment.
Initiating Events Analysis in Operating Phase for Pressurized Water Reactor Nuclear Power Plant
Authors: Chu-ran Feng , Yuan Ma , Lin Yan , Guansan Yang , Bin Liu
Published: November 2022

Probabilistic Safety Assessment (PSA) is an important tool for risk-informed integrated decision making application. Initiating events are caused by equipment failure or/and personnel error which causes disturbance to the nuclear power plant (NPP) and may lead to core damage. As the start of PSA, the quality of initiating events analysis has an important influence on safety assessment and PSA application for NPPs, and the application of operating experience of domestic and aboard are of great importance in initiating events analysis. Based on analysis experience of PSA model in design phase and operation phase for different NPPs, this paper developed a set of method for initiating events analysis of operation phase PSA model development. Combined with the referenced initiating events list, engineering evaluation, operating experience analysis and main logic diagram, a relatively complete initiating events list have been determined. According to the characteristics of the initiating events and analysis requirements for PSA applications, this paper adopted statistical analysis method or specific analysis method to determine the frequency of initiating events which could reflect the design of specific NPPs and operation level of domestic NPPs. The development and application of these initiating events analysis method, which could reflect design and operating characteristics more realistically, improved the development level of PSA model in operation phase effectively, and laid a solid foundation for PSA applications.
Investigation on Methodology of Coastdown Analysis and Design for Reactor Coolant Pump
Authors: Zhong Yun
Published: November 2022

Methodology of coastdown analysis for RCPs was investigated with consideration of combined effect of loop fluid inertia and pump inertia. A simplified methodology based on momentum equations of the loop and the pump respectively was developed to describe the combined effect, leading to a different guess of flow coastdown rather than speed coastdown. Both the traditional axial-sealed RCP and a canned motor sealless RCP experiment data on test rigs were used to verify the accuracy of the abovementioned methodology for coastdown flow calculation. It has been verified that the calculated coastdown flow fits with the measured data better when considering the influence of loop fluid inertia. Then a coastdown performance oriented design method considering the influence of loop fluid inertia was developed based on coastdown analysis methodology investigation, which can increase pump inertia flexibility without violation of nuclear safety and benefit for balance of safety, economy and technical feasibility. Using the slope of 1/q coastdown curve, which is approximately a straight line, makes RCP inertia determination much more definitely. And the methodology of coastdown analysis and design can be applied to sealless RCP when using overall efficiency without electric loss as input. Especially, deduction of electrical loss from the overall efficiency is important for canned motor RCP, because it has significantly large electrical loss due to stator can, while this loss is not applicable during coastdown without electric power supply.
The Research and Application of Grouped Control in Airborne Radioactive Containment Ventilation Systems
Authors: Du Pengyu , Tang Shuang , Zeng Xianbin , Peng YongSen
Published: November 2022

The application of the grouped control technology can significantly simplify the interface of the human-machine; improve the automatic control level and the efficiency of the operators.

A typical grouped control scheme of the airborne radioactive containment ventilation systems is analyzed in this paper based on the application in the HPR1000 nuclear power plant projects constructed in China. And the characteristics of the grouped control scheme are studied and summarized.
A Collaborative Edge Computing Framework for Predictive Maintenance in Nuclear Power Plants
Authors: Feng Yixiong , Wang Yong , Hu Bingtao , Hong Zhaoxi , Tan Jianrong
Published: November 2022

Nuclear power is an indispensable part of modern energy systems. To operate the nuclear power plants safely and reliably, it is crucial to greatly develop the predictive maintenance of nuclear infrastructure with the support of various smart sensors and big data analytics. To this end, this paper proposes a novel collaborative edge computing-enabled solution for predictive maintenance in nuclear power plants, from which a key problem of efficiently allocating some edge computing tasks is formulated. Specifically, considering huge amounts of industrial data are continuously generated during plant operations, we first present a three-layer predictive maintenance computing framework for nuclear power plants. Subsequently, to timely process these data in some distributed and heterogeneous industrial computing nodes, a complicated scheduling optimization model with some interdependent computational tasks is established. To lower the size of model, we also introduce some reduction strategies. Finally, an actual predictive maintenance scenario in nuclear power plant is chosen and some algorithms are taken for comparisons.
Disorders in Fluid Filled Pipeline Structure With Elastic Helmholtz Resonators
Authors: Zeng Qingna , Wang Donghui , Zang Fenggang , Zhang Yixiong , Wang Bihao , Yuan Zhihao
Published: November 2022

This paper studies the influence of hybrid disorders on acoustic transmission characteristic in pipeline with periodic elastic HRs. Band gap structures (BGs) for infinite period are investigated by transfer matrix method and finite element method respectively, and fit well with the corresponding transmission loss (TL) for finite regular periods. By introducing a certain degree of uncertainty level, Interval analysis method is used to convert uncertain parameter problems into two deterministic models. Frequency response function (FRF) for near-periodic structure with single disorder and BGs for quasi-periodic structure with multiple disorders considering cavity radius and periodic length are investigated. The existence of disorders in periodic structure would always reduce attenuation interval and intensity, some disturbance would even change the original acoustic transmission characteristic and excite completely new attenuation zones. The sensitivity of pipeline acoustic vibration transfer characteristic to disorder parameter depends on attenuation mechanism, as for periodic elastic HRs pipeline, acoustic propagation stability is more sensitive to characteristic of locally resonant unit. The work in this paper enriches the stability analysis of pipeline with elastic HRs, which would provide theoretical reference for noise and vibration reduction design in practical engineering applications.
Efficient Dynamic Monte Carlo Simulation in Multi-Physics Coupling Calculation
Authors: ShangGuan Danhua , Yan Weihua , Wei Junxia , Gao Zhiming , Chen Yibing , Ji Zhicheng
Published: November 2022

If particle transport is included in some multi-physics calculation, Monte Carlo method is often used to simulate it and occupies the largest amount of calculation time. So, efficient dynamic Monte Carlo simulation for time-dependent particle transport problem is important, which is inevitably relied on large-scale parallel calculation. Two methods are proposed in this paper. The one is a tally-reduce algorithm which is used in the coupling of transport simulation and burn-up calculation. By reduces the amount of data which should be reduced necessary, this method can decrease the tally-reduce time largely. It can be seen as a new coupling mode for these two processes in MPI environment and will has larger value in cases when model scale is larger relatively compared to sample size. The other method is an adaptive method for setting the sample size of Monte Carlo simulation. Relying on the generalization of the Shannon entropy concept and an on-the-fly diagnosis rule for a entropy value sequence, the adaptive method proposed in this paper can decrease the original huge sample scale to a reasonable level. By numerical test for some non-trivial examples, both algorithms can decrease the calculation time largely while make the results almost unchanged.