bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.