bluuebunny/crossref_metadata_2025 huggingface dataset:
Radiation Shielding Towards Commonly Available Objects
Authors: Zamani Ahmad Muhammad Amir Safwan , Sazali Muhammad Arif , Ripin Azuhar
Published: November 2022

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

At present, the technology of reactor neutron noise is developing rapidly, which can not only be used to measure some intrinsic parameters of the reactor core, but also be used for core fault diagnosis. The theoretical solution of reactor neutron noise is of great significance for providing information on the directivity of core faults, and its analytical solution is extremely difficult or even impossible. Therefore, it is necessary to study the numerical simulation method of reactor neutron noise. In this work, the governing equations of the neutron noise problem are firstly established based on the discrete ordinate finite element theory, the equations take the reactor steady-state parameters and the core disturbance cross-section as input parameters. Secondly, the frequency domain solver FrequencyDomain_SNSolver of neutron noise is developed, which is capable of acquiring steady-state parameters, reading disturbance cross-section and calculating neutron noise. Finally, for a calculation case of an experimental reactor core, the neutron noise distribution and spectrum under perturbation are obtained by using the developed FrequencyDomain_SNSolver. This work can provide some reference and technical support for related research.
Seismic Analysis of the Scale Reactor Structure Model of a Lead-Bismuth-Cooled ADS Reactor
Authors: Rong Gang , Zhu He , Yang Yi
Published: November 2022

This paper is aimed to analyze the structural responses of a scale reactor structure model of a lead-bismuth-cooled ADS reactor under the design seismic time history loads. As a heavy-liquid-metal-cooled reactor, the fluid-structure interaction due to seismic loading is a key issue that may significantly affect the responses of the structures. Two three-dimensional finite element analysis models based mainly on shell element have been built, considering the gap and contact effect between components by nonlinear contact elements with gap specification. The fluid-structure interaction of the structure with an annular cavity in the scale reactor structure was included in the analysis model through a hydrodynamic mass matrix element considering no leakage effect on the ends of the annulus for the first model, and assuming some mass loss due to leakage at the ends for the second model, while additional mass elements due to fluid effect were used for all other structures in both models. Simulation analyses for the scale reactor structure were carried out to obtain the dynamic acceleration response data of the main components, and the comparison with measured data of vibration characteristic tests was conducted, the result showed that considering some mass loss at the ends of annulus would obtain a more precise result. The results of this paper will provide a theoretical recommendation for the design scheme of the subsequent seismic experiment of the scale reactor structure model.
Development and Validation of a New Gas Conductance Model
Authors: Yang Shengzhi , Wei Xiaoyan , Zhu Yanan
Published: November 2022

The analysis of fuel rod behavior has become increasingly important and reliable assessment of temperature distribution is vital. Gap conductance is a key parameter to calculate the heat transfer between fuel pellet and cladding, then fuel temperature and thermo-mechanical of fuel rods will be influenced. Gas conductance often plays the most important role comparing convective heat transfer, thermal radiation and solid contact conductance in fuel rod analysis. A new gas conductance model has been proposed, considering the effects of mixture gaseous component, the temperature discontinuity on the surfaces of pellet and cladding, and the roughness of solid surface, since the irregularity will become significant when the gap is closed at the extended burn-up. Moreover, the model has been calibrated with Modified Pulse Design (MPD) and Modified Longitudinal Design (MLD) experiment gas conductance data performed by John. E. Garnier and Stefan. Begej, and shown a good prediction accuracy at various experiment conditions. Furthermore, the model has been successfully implemented in fuel rod performance analysis code JASMINE to improving the temperature distribution prediction. Based on the comparison of code prediction of centerline temperature with HALDEN in-core irradiation tests data, the results show that the new gas conductance model has a decent suitability to JASMINE.
Numerical Investigation of Melt Fragmentation in Sodium-Cooled Fast Reactor Based on Moving Particle Method
Authors: Wang Wenpeng , Zhang Bin , Cao Sheng , Hu Minghao , Shan Jianqiang
Published: November 2022

The Melt-Fuel-Coolant Interaction (MFCI) is one of the important phenomena during reactor severe accidents. In particular, the fragmentation process of the molten fuel in the metal coolant is a complicated multi-phase and multi-component heat transfer phenomenon accompanied by solid-liquid coupling and phase change, which is of great significance to the formation of debris bed on core catcher as well as the coolablity of debris bed after severe accident. It is of great significance to conduct research on MFCI of sodium-cooled fast reactor. A moving-particle-method based program is developed for the melt fragmentation analysis of sodium-cooled fast reactor in the present study. The program is based on the finite volume particle method and improved free surface pressure model, smoothing model, heat transfer and phase change model, interface surface tension model and solid-liquid coupling model are incorporated. Afterwards, the process of fragmentation is compared against the experimental data. The effects of initial temperature of melt, velocity of melt entering coolant, and jet diameter on the fragmentation are analyzed. The results show that the initial temperature of the melt has a great influence on the melt fragmentation, and a higher initial temperature makes the melt fragmentation more complete. Secondly, in the simulation of melt jet fragmentation, the process can be divided into six phases by analyzing the variation of the amount, the maximum mass, and the maximum size of fragments with time. Thirdly, by changing the initial velocity and diameter of the melt jet, it is concluded that the higher initial velocity and the larger melt jet diameter will produce a higher peak amount of fragments.
Probability Safety Analysis for Nuclear Power Plant Long-Term Temporary Shutdown State
Authors: Chen Yanzhu , Yang Zhi
Published: November 2022

In recent years, due to the influence of power grid demand and other factors, some nuclear power plants have long-term temporary SHUTDOWN. At present, there is no mature quantitative analysis method to measure which Plant operational state is safer under long-term temporary shutdown. Taking a second-generation nuclear power plant as an example, this paper first briefly introduces the PSA analysis method; Secondly, the difference is analyzed qualitatively, which between standard long-term temporary shutdown condition and refueling shutdown condition. Thirdly, from the quantitative point of view, the model processing methods and analysis assumptions are introduced, the risk views are given, and the quantitative calculation and sensitivity analysis are carried out; Finally, the analysis conclusion is obtained.

Stay in the standard long-term temporary shutdown state for 15/30/45/60 days, after supplementing the calculation of the thermal shutdown condition,the risk per unit time of long-term temporary staying Maintenance Cold Shutdown mode is lower. The risk sequence is: Cooldown with the residual Heat Removal mode > Hot shutdown condition > Maintenance Cold Shutdown mode with closed primary loop. From the perspective of the risk of core damage, it is suggested that the Maintenance cold shut mode in which the primary loop is kept closed is recommended for long-term temporary shutdown of the power plant.
Layout Optimization of Self-Powered Neutron Detector for HPR1000
Authors: Zhou Yao , Cao Liangzhi , Zhao Changyou , Li Zhifeng , Zhu Yuxiang
Published: November 2022

Self-Powered Neutron Detector (SPND) has been widely applied for the on-line monitoring of the key core parameters, including the power distribution, power-peak factor and so on. For the SPND layout design, the vital parameters include the SPND amount and loading position in the radial direction and segment amount in the axial direction. Different SPND layout reflects different ability of power distribution reconstruction, which is directly related to the safety and economy of nuclear power plant. For HPR1000, the neutronics sensitivity of SPND layout design is analyzed based on a 3D on-line power distribution monitoring system NECP-ONION. Through the comparison and analysis of the reconstruction errors of power distribution, the design rationality of SPND is evaluated and the optimization design is recommended. For the recommended SPND layout design, the loading positions in the radial direction are arranged at intervals and segment amount in the axial direction is 15. The recommended SPND layout design shows stable performance in the whole life for HPR1000 reactor, and the RMS errors of the power distribution reconstruction has been in the range of [0.32%, 0.48%], meeting the safety limit and accuracy requirements of power distribution reconstruction. This work verifies the on-line monitoring ability of NECP-ONION system. Additionally, it provides a technical reference for the SPND experiment in HPR1000.
Research on Martensitic Stainless Steel Used on the Latch Housing of European Pressurized Reactor (EPR) Control Rod Drive Mechanism (CRDM)
Authors: He Cuizhu , Lan Yinhui
Published: November 2022

As one of the most important equipment for reactivity control, Control Rod Drive Mechanism (CRDM), which is widely used in Pressurized Water Reactor (PWR) nuclear power plant, has a series of important security functions. As part of the reactor pressure boundary, materials for European Pressurized Reactor (EPR) CRDM pressure housing needs to have relatively high strength and sufficient toughness (i.e., appropriate yield ratio). As an important part of the magnetic circuit, martensitic stainless steel used for latch housing (a part of pressure housing) needs to have good magnetic properties. It is difficult for the yield ratio and the magnetic properties of the materials to meet the specification requirements simultaneously. Based on the design specification, this research refers to the relevant international manufacturing experience and combines the current manufacture with engineering practice. The manufacturing process and key parameters of the material for latching housing of CRDM are established. The martensitic stainless steel which meets the design requirements has been successfully prepared. The comprehensive properties (e.g. yield ratio and magnetic properties) of the material are improved as well. The performance indicators have reached international advanced level. The successful development of the martensitic steel for EPR CRDM latch housing provides strong technical support for the construction of nuclear power projects at home and abroad.
Research and Practice on Operational Risk Management of Nuclear Power Plants
Authors: Wei Haifeng
Published: November 2022

Risk management is a very important work for nuclear power plants, which is related to the sustainable, healthy and stable operation of nuclear power plants. Almost all the work of nuclear power plants requires risk identification and risk control. Based on the risk management theory, advanced concepts and good practices, this paper defines the operational risk management of nuclear power plants, puts forward five modes of operational risk management of nuclear power plants and corresponding control measures for the first time. Combined with the practical feedback of nuclear power plant operation, it clearly puts forward the principles and suggestions for formulating operational risk control and response measures of nuclear power plants. Referring to the analysis mode and corresponding control measures of nuclear power plant operational risk management proposed in this paper, we can accurately and comprehensively analyze the operational risk of nuclear power plants, formulate the risk control measures of validity, the effectiveness of operational risk management of nuclear power plants in guaranty, and prevent the occurrence of accidents/events. At the same time, the research results of operational risk management of nuclear power plants have important reference and guiding significance for the promotion and optimization of risk management of nuclear power plants.
Study on the Effect of Random Distribution of the Fuel in Dissolver
Authors: Yu Miao , Yi Xuan , Shao Zeng , Yang Haifeng
Published: November 2022

The main role of the dissolver is to dissolve sheared short sections of the spent fuel assembly. During the dissolution process, there is the situation where solid fuel coexists with and randomly distributed in the solution. Generally, the optimal moderation grid array is used as the bounding model in the criticality safety analysis, which has a considerable conservative margin. Based on a hypothetic dissolver equipment, this paper studies the effect of the random distribution of the solid fuel in the dissolver. Sensitivity analysis is carried out on the random distribution states, fuel composition, solid fuel volume proportion and radii of fuel particles. According to the results of sensitivity analysis, the suggestion for criticality safety analysis model of dissolver is given. The research on the effect of the random distribution of the solid fuel in the dissolver provides a reference for the optimization of the criticality safety analysis model of the dissolver, which can be referred in the subsequent criticality safety analysis of the dissolver to increase the processing capacity of the dissolver and improve the dissolver economy on the premise of ensuring safety.
Multiphysics Thermo-Mechanical Behavior Modeling for Annular Uranium-Plutonium Mixed Oxide Fuels in a Lead-Cooled Fast Reactor
Authors: Cai Yutai , Zhou Wenzhong , Gu Long
Published: November 2022

A model for fuel performance of annular uranium-plutonium mixed oxide (MOX) fuels in lead-cooled fast reactors is developed. This model has several modules working in fully coupled approach based on COMSOL Multiphysics. These modules describe (1) Fission gas release, (2) Plutonium redistribution, (3) Oxygen redistribution, (4)Heat transfer, (5) Gap width evolution. The modeling results were benchmarked with existing codes and experiment data, the results were in satisfactory agreement. The parameters analysis was carried out in this work. Consistent with the solid MOX fuel, the O/M ratio (or deviation from stoichiometry of oxygen) has a great effect on temperature evolution, fission gas migration and release behavior and plutonium redistribution. The effect of linear power on fuel performance are also analyzed.
Assessment of Passive Cooling Capability of the Westinghouse Lead Fast Reactor Under Station Blackout Conditions
Authors: Tatli Emre , Lee Sung Jin , Liao Jun , Ferroni Paolo
Published: November 2022

Westinghouse is developing its next generation of high-capacity nuclear power plants with the Westinghouse Lead-cooled Fast Reactor (LFR) (Ref. 1). The Westinghouse LFR is a pool-type 950 MWt (∼450 MWe Net) passively safe modular construction plant that contains all the primary system components within the reactor vessel, including the reactor coolant pumps (RCPs) and the primary heat exchangers (PHEs). Similar to the Westinghouse AP1000® Pressurized Water Reactor design, the Westinghouse LFR design incorporates passive cooling capability primarily through its Passive Heat Removal System (PHRS). During accidents, the PHRS allows decay heat to be transferred from the core to a pool of water surrounding the guard vessel, which transitions to long-term air cooling upon water depletion. All of this takes place without operator action or need for moving parts or power supply. Overall, when combined with natural circulation of liquid lead inside of the reactor vessel, this passive system fully relies on natural circulation, thermal radiation and boiling mechanisms to transport decay heat from the core to the atmosphere. In this study, complementary Computational Fluid Dynamics (CFD) and system code analyses were performed to assess the passive cooling ability of the Westinghouse LFR design during station blackout (SBO) conditions. The postulated scenario examined in this study is the longer-term phase of the SBO event, with the RCPs and PHEs assumed to be inactive and the water inventory in the pool exterior to the guard vessel fully depleted, thus relying purely on air cooling as the ultimate heat sink. The CFD analysis was conducted with Siemens STAR-CCM+ code using, as domain, the primary lead pool and with boundary conditions provided by the containment code GOTHIC. The CFD analyses provided high fidelity resolution of the three-dimensional natural circulation flow patterns formed throughout the reactor vessel. The results of the study not only provided confirmation of the passive cooling capability of the Westinghouse LFR design during an SBO event, but also provided granular input for vessel structural analysis activities accounting for thermal gradients developed during an SBO event.
Study on Hydrogen Risk Venting Mitigation Measure of Small Steel Containment During Severe Accident
Authors: Zhiqiang Zou , Hang Zhang , Ming Zhang , Huanhuan Peng , Youyou Xu , Jian Deng
Published: November 2022

The hydrogen risk in small steel containment is one of the most important safety problems of small modular reactor (SMR), effective measures need to be taken to mitigate the risk of high concentration hydrogen in a small free volume during severe accidents. In the early stage of a severe accident caused by a loss of coolant accident, the pressure in the containment may rise rapidly, probably triggering the immediate venting of the containment, which can protect the integrity of the containment in the stage of pressure rise in the short term. In the later stage of a severe accident, the generation of hydrogen from reactor core may cause hydrogen combustion or even explosion, which may cause the possibility of overpressure in the containment. In this paper, Computational Fluid Dynamics (CFD) GASFLOW-MPI code is used to analyze the effect of containment venting measures on hydrogen risk in small steel containment of SMR during severe accident initiated by direct vessel injection line break loss-of-coolant. The analysis results show that the venting action can effectively depressurize the containment, the larger the area of the venting pipe is, the more obvious the depressurization effect is. However, hydrogen cannot be effectively discharged in the middle and later stage of the accident, so the risk of hydrogen cannot be completely eliminated, and the possibility of flame acceleration in containment cannot be ruled out for the reason that the negative pressure generated by water vapor condensation will suck in the air outside the containment and form a combustible gas cloud again.
Prediction of Critical Heat Flux Based on Random Forest Algorithm
Authors: Sun Zhuo , Han Mingyang , Yuan Peng , Gong Lei , Chen Di , Yuan Guanghui , Zhang Shuang
Published: November 2022

The accurate prediction of critical heat flux (CHF) is of great significance to the safe operation of the reactor. At present, most of the traditional methods for predicting CHF are based on relational fitting in mathematical form, which are not accurate enough and may can not taking full advantage of experimental data. Therefore, a prediction model based on random forest algorithm (RF) is proposed, the model is based on CHF 2006 look-up table, and the accuracy and smoothness of the model have been verified when predicting CHF. In this paper, the variation trend of CHF predicted value and actual value with pressure, mass flow rate and thermodynamic quality is given. It has been verified that the error of this method is acceptable, the predicted value is in good agreement with the experimental value, and conforms to the corresponding physical laws, which can provide a reference for the development of high-precision CHF prediction models or CHF-related experiments.
Optimization in Heat Transfer Scheme for Load Rejection and Emergency Shutdown Conditions of Sodium Fast Reactor
Authors: Yu Yang , Yang Jun , Wang Jin , Wang Xiaokun , Hu Wenjun , Zhang Donghui
Published: November 2022

For a demonstration pool-type sodium fast reactor (CFR600), load rejection and emergency shutdown conditions put forward sharp requirements for the reactor control system by introducing abrupt changes in power and flowrates. While the sodium pool has a slow temperature response resulting from the heat capacity, the change in the circuit flowrates at the initial stage of the transients could lead to fluctuations in the intermediate heat exchanger, steam generator, and other essential devices. This research firstly clarified the constraints and objectives of the working condition design, and adopted an envelope calculation method to optimize the scheme under load rejection and emergency shutdown conditions. The research proposed several optimizations of secondary pump flowrate and the BOP water supply scheme, which enhanced the stability of devices in transient conditions. The validity of the scheme is finally re-checked from the perspective of energy transportation. It is concluded that by targeting at increasing heat transfer and reducing equipment fluctuation in the initial stage of transient, the optimized scheme realizes the balance between safety and economy to a certain extent. This research may provide the guidance for experiment configuration and the further study.
Rod Ejection Accident Simulation of a Small Pressurized Water Reactor
Authors: Fan Ziqi , Sun Peiwei , Wei Xinyu
Published: November 2022

Rod Ejection Accident (REA) is one of the major accidents which may result in the serious consequences during nuclear reactor operation. The study on REA is helpful to understand the physical characteristics and to verify the safety design of the core. The small pressurized water reactor (SPWR) adopts a relatively small amount of control rods. Therefore, when REA occurs, the positive reactivity introduced by the accident may be larger, resulting in more serious consequences. The integral design of SPWR results in different responses to the REA. Thus, it is necessary to investigate the REA of SPWR.

In this paper, 3KeyMaster simulation platform, which is coupled with a thermo-hydraulic model and a three-dimensional space-time neutron dynamics model, is used to establish the reactor core model. Four single-beam control rods are selected to carry out REA, respectively. Simulations are performed, and results are collected. The responses of the SPWR are analyzed and accident procedures are obtained.

Through analysis, it can be concluded that the serious consequences are caused when the R rod closest to the core, resulting in the maximum nuclear power peak. But the fuel center temperature and cladding temperature do not exceed the limit value, which proves the safety of SPWR and the rationality of core design.
Analysis of Hidden Fault in Digital Instrumentation and Control System of Nuclear Power Plant Based on Environmental Model
Authors: Guo Chao , Zhou Shuqiao , Li Duo , Huang Xiaojin
Published: November 2022

Nuclear safety is one of the key issues for a nuclear power plant (NPP). The instrumentation and control system (I&C system) plays critical roles for the safe and efficient operation of an NPP. Due to the large scale and complex logic of the I&C system, some hidden faults cannot be found by factory testing or on-site debugging and may cause system degradation or even system failure during NPP operation. These hidden faults may come from system design omissions, software programming errors, personnel operation failures, natural environment, inappropriate preventive maintenance, etc. Timely identification of these hidden faults is of great significance for improving the reliability of the I&C system. This paper takes the hidden faults of the I&C system as the research object, and carries out the analysis and investigation of the hidden faults based on the environmental model. After determining the key and important characteristics of the I&C system to be analyzed, this paper analyzes and classifies the triggering causes of the faults from three dimensions of specific application environment, specific application method and long-term dynamic operation. This paper takes the hidden fault “Failure of Power Module of the DCS Cabinet”, “Operation failure of KCS troubleshooting personnel”, and “Communication abnormal of DCS internal network” actually found in the I&C system of an NPP as cases to establish the mapping relationship between the hidden fault and the environmental model to verify the correctness of the built model. The work showed in this paper can contribute to improve the identification capability of hidden faults in I&C system of the NPPs, which has important engineering significance for improving the reliability of the I&C system.
Research on Electromagnetic Interference Protection of Nuclear Power Unit Equipment Under the Application of New Wireless Technology in Smart Nuclear Power
Authors: Chen Bing , Zhao Guidan , Zhong Zhiping , Zheng Yong , Yu Charlie
Published: November 2022

Intelligence is the trend of industrial development in today’s world, which will inevitably affect the nuclear power field, thus promoting nuclear power industry from digital to intelligent forward.5G and Wi-Fi6, as a new generation of wireless communication technologies, have important advantages in terms of technical performance and security due to the characteristics of dense personnel, wide equipment distribution, complex facility environment and special materials used in nuclear power scenarios. Considering the high requirement of electromagnetic compatibility in nuclear power environment, it is urgent to further study the electromagnetic interference and protection requirements of 5G, Wi-Fi6 and other new technologies on nuclear power unit equipment in nuclear power environment.

In this paper, combined with relevant national document standards and industry standards, we will sort out the requirements of Immunity about Radio Frequency Radiation (RFR) for the important equipment in nuclear power plants, and use signal generators to simulate new radio signals such as 5G and Wi-Fi6 to test the immunity of nuclear power instruments and meters. Combined the results of test, we will study the Electro Magnetic Compatibility (EMC) between new wireless technologies and nuclear power plant equipment from different factors, such as Power, bandwidth, modulation, distance, direction. And then, we will summarize the protection requirements about nuclear power equipment.
Research and Application of Automatic Modeling Technology for Mechanical Analysis of Nuclear Power Pipelines Based on PepS
Authors: Gong Xiao Cheng , Jin Xian Jing , Zhang Yu
Published: November 2022

Mechanical analysis of nuclear power pipeline is an important part of nuclear power design. At present, the widely used nuclear-grade pipeline mechanical analysis software in the world is PepS, which almost contains the commonly used standards and specifications in the nuclear power industry. PepS software is used to carry out the mechanical analysis of nuclear-grade pipeline. The key step is building a calculation input pipeline model that accounts for more than 70% of the entire pipe mechanical analysis workload. At the same time, in order to meet the strict process layout requirements of nuclear power, a pipeline usually needs to undergo multiple mechanical analysis and demonstration. There are thousands of pipelines in nuclear power projects, and its mechanical analysis needs to invest a lot of manpower cost. This research deeply analyzes the procedures and principles of PepS calculation input model, the topological structure and geometric characteristics of the 3D model of the pipeline based on the three-dimensional layout design results, and realizes the automatic modeling technology of mechanical calculation input model of nuclear power pipeline based on PepS software through digital and intelligent information technology. The implementation of this technology completely replaces the manual modeling, eliminates the human error of manual modeling, fundamentally liberates the productivity, obtains remarkable economic benefits, and helps the nuclear power design company go hand in hand with multiple projects to realize batch quantitative design and production.
Fault Diagnosis of CRDM Based on EEMD-Wavelet Threshold Denoising and WOA-KELM
Authors: Jingdong Luo , Ling Chen , Liming Zhang
Published: November 2022

CRDM (Control Rod Drive Mechanism) is the key equipment of nuclear reactors. Accurate fault diagnosis can improve the reliability of CRDM equipment and effectively and quickly eliminate related faults, thus ensuring the safety of reactors. However, due to its complex operating environment, the fault characteristics of CRDM are easily masked by complex and changeable noises, resulting in difficulty in fault identification. To solve this problem, this paper proposes a fault diagnosis method for KELM based on ensemble empirical Mode decomposition (EEMD), wavelet threshold denoising and whale optimization (WOA). Firstly, an EEMD-Wavelet threshold signal denoising method is proposed according to correlation criterion and kurtosis criterion. Then, the time domain feature, frequency domain feature, EEMD modal component energy and sample entropy feature of noise reduction signal were extracted to construct a multi-scale feature set that can accurately reflect the fault state. Finally, a classification method based on WOA-KELM is proposed to diagnose the fault feature set accurately. The proposed method is validated with CRDM vibration signal data set, and compared with common fault diagnosis methods, the results show that the fault diagnosis method based on EEMD-Wavelet threshold denoising and WOA-KELM has higher fault classification accuracy.
A MIN-TPN-Based Remote Disaster Recovery and Backup Scheme of Power Industrial Control System in Nuclear Power Plants
Authors: Gang Cui , Yang Xin , Li Hui , Wang Han
Published: November 2022

Nuclear power plants are important industrial production facilities whose safety issues can have disastrous consequences for the social environment. With the deep fusion of industrialization and information, the nuclear power control system is gradually transformed into automatic digital systems. Such development improves the production efficiency and economic benefits, but introduce the security problem from traditional IT systems into industrial control systems. However, the existing power industrial control system of nuclear power plant mainly adopts firewall, anti-virus software, network gate and other passive security strategies, which lack the safety of active. Besides, only part of the data can be backed up locally and the remote backup is completed manually by operators regularly. The security of industrial control systems cannot be guaranteed and real-time remote backup cannot be realized. In this paper, we analyze and summarize the characteristics of industrial control systems used in nuclear power plants. Then we propose a Truly Private Network (TPN) for nuclear power plants based on the Multi-identifier network (MIN), which integrates blockchain, cryptography, and other security mechanisms to provide a trusted environment. Lastly, we build a network remote real-time backup scheme based on Mimic distributed storage technology providing data tamper-resistance and traceability for nuclear power plants.
The Research on Welding Procedure and Performance of Zr-3 Thick Plate
Authors: Gao Yu , Liang Xiaoru , Wang Yuxin , Guo Lifeng
Published: November 2022

Due to the excellent corrosion resistance and small absorption interface of thermal neutrons, Zr has been widely used in nuclear industry. As an important manufacturing process, the welding of Zr is usually performed by manual tungsten inert-gas arc welding (TIG). Due to the low deposition efficiency of tungsten inert-gas arc welding (TIG), it is more suitable for thin plates. If this method is used for thick plates, it will increase the cost of time, manpower and material resources. Therefore, for thick plates, an advanced and efficient welding method is required to improve the welding quality and construction efficiency. In this paper, a combination of plasma arc welding (PAW) for root bead and tungsten inert-gas arc welding (TIG) for filling and cap beads was successfully developed for 20mm Zr-3 plate, with specifically designed clamp and gas shielding devices based on the welding characteristics of Zr. Different post-weld heat treatment was performed for the weld pieces. Moreover, the mechanical properties and corrosion resistance of welded joints were evaluated. The results demonstrates that welded joints which meet the requirements specified in code and standard and engineering requirements can be obtained by developed welding process and welding system; the influence of different heat treatment on mechanical properties and corrosion resistance of welded joints was not obvious.
Analysis Method and Application of Maintenance Rules for Main Steam System in Qinshan no.1 Nuclear Power Plant
Authors: Jin Yan Jun
Published: November 2022

Nuclear power plants have traditionally adopted regular preventive and corrective maintenance methods, but extensive international operation experience shows that inadequate maintenance, excessive maintenance and improper maintenance will adversely affect nuclear power plant structures, systems and equipment, and thus affect the safety level of nuclear power plant operation. According to China’s National Nuclear Safety Administration of the improved technology policy (try out) “the effectiveness of the nuclear power plant maintenance requirements, with the first nuclear power qinshan owner steam system as an example, the implementation based on the structures, systems and equipment effectiveness analysis method and risk management of the maintenance rules, can be more reasonable and effective to improve the reliability of maintenance.

The method mainly includes: determining the system boundary and the scope of structures, systems and equipment involved in the maintenance rules, screening the system maintenance rules function; List the equipment and failure modes that affect the function; The failure probability and function importance of each function group are calculated. The reliability index and availability index of the system are determined and verified by historical data. Supervise each function group of the system according to the key indicators, evaluate the actual operation of these structures, systems and equipment regularly, determine whether to meet the established performance indicators, optimize and adjust the maintenance rules according to the evaluation results, so as to avoid insufficient and excessive maintenance.
Development and Application of High Temperature and Radiation Protection Equipment for Spent Fuel Dry Storage System
Authors: Cheng Cheng , Yuanyuan Wei , Yongjie Pan , Yongjun Ge , Zhiguo Wei , Zhihui Zhou , Zengfen Li
Published: November 2022

Based on the safety analysis, the risk of high temperature and radiation are found in the local operation during the automatic welding and vacuum drying of typical spent fuel dry storage system. To enhance the operational safety, a high temperature and radiation protection equipment has been developed through the structural design and numerical analysis and has been applied in demonstration project in China. It is proved by the local monitoring data that in the case of loading spent fuels with a total heat rate of 26kW, with the help of the protection equipment, the highest temperature of the operational area is 60% lower than the case without the equipment in the same environmental conditions. The dose rate in the area can also be 60% lower. The equipment is proved to be of significant help to ensure the safety of operators in high temperature and radiation environments, during both normal operation and equipment failure conditions.
Analysis and Improvement of Flow Field Calculation in Thermal Hydraulic Program for High Temperature Gas-Cooled Reactor
Authors: Haoming Dou , Minggang Lang , Han Zhang , Ding She
Published: November 2022

The flow field calculation plays a crucial role in thermal-hydraulic calculation of High Temperature Gas-cooled Reactor (HTGR). While using the thermal-hydraulic program for HTGR to calculate the low-flow condition, the calculation of the helium pressure field distribution in the reactor core is difficult to converge, which may lead to the non-physical pressure oscillation. The analysis and improvement of flow field calculation to alleviate this non-physical oscillation is of great significance for HTGR thermal-hydraulic design. In this work, the factors affecting the flow field calculation are analyzed. The approximation assumptions of horizontal cavity modeling and the using of Gauss-Seidel (GS) partition point iteration method leading to the flow field calculation is sensitive to round-off error. For the reason of GS partition point iterative method suffers from the slow convergence rate, and may even non-convergence in some conditions, a global direct method is developed in this paper to solve the whole flow field directly to pursue strong stability and high convergence rate. Compared with original method, the numerical results show that the newly developed algorithm achieves faster convergence rate and stricter convergence results.
Research and Application on Drawing Method of Optimizing for the Intelligent Process System Flow Chart of the Nuclear Power
Authors: Chen Haiyan , Tian Licheng , Zhou Zhigang , Yang Jiaojiao
Published: November 2022

The intelligent process system flow chart of the nuclear power is based on the symbol library. Each component and its attributes on the system flow chart are stored in a digital form for reference to subsequent projects. The content of this paper is the automatic generation of Definitive Data Assembly and the automatic association of the drawing interface of the intelligent process system flow chart by using Diagrams. It is an optimization of the intelligent process system flow chart for the application of Diagrams by using information technology and computer language. The Definitive Data Assembly automatic generation function realizes generating Definitive Data Assembly statistical table at a fixed position of the drawing surface, and improves the neat specification of the drawing information; The single drawing Definitive Data Assembly statistical table is automatically generated, and the batch drawing Definitive Data Assembly statistical table can be automatically generated, one-click operation, which greatly improves the work efficiency; at the same time, solves the problem of disordered management of Definitive Data Assembly. Optimized the application of Diagrams to draw the interface merging function of the intelligent process system flow chart, which greatly reduced the number of files in the project file XXXMAC->Schematic path, reduced the selection obstacles and improved the correct rate; Developed interface-connected menus, one-click operation, saving time for drafters, automatically deleting MAC files, reducing the workload of project administrators, and releasing project storage space, improving work efficiency. The research results of this paper are applied to draw the process system flow chart of many specialities such as nuclear island system, radioactive waste management, hydraulic technology, HVAC, chemical water, conventional island machinery, etc.
Analysis of EBR-II Loss-of-Flow Transients With 3D Model of CATHARE Code
Authors: Chen Xiaolin , Li Simon , Cheng Songbai
Published: November 2022

As one of the GEN-IV nuclear reactor systems proposed by GEN-IV International Forum, the sodium-cooled fast reactor features the energy generation with fast spectrum and a closed fuel cycle for fuel breeding and actinide management. In order to validate system codes for reactor system analyses and to improve code performance, benchmark activities of experimental transients of fast reactors were performed by worldwide institutions. The Shut-down Heat Removal Tests in Experimental Breeder Reactor II were proposed in the context of the International Atomic Energy Agency. In this work, loss-of-flow tests performed in Experimental Breeder Reactor II are analyzed using system thermal-hydraulic code CATHARE. Evolutions of important reactor parameters such as core power, coolant flow rate, temperatures in core and in the intermediate heat exchanger are predicted and compared against experimental data. The sodium pool is further modeled by using a 3D model in CATHARE-3 and its verification is also conducted in the benchmark. Qualitative agreement is obtained for the reactor parameters predicted by CATHARE. The inherent passive safety characteristics of Experimental Breeder Reactor II in unprotected loss-of-flow transient is demonstrated, and the stratification effects in sodium pool are able to be identified with the 3D model. The results obtained in this work are expectable to provide some valuable insights for future sodium-cooled fast reactor transient accident analyses by coupling system thermal-hydraulic codes and CFD tools.
Optimization the Protection Configuration of the RAM Excitation System of the Nuclear Power Plant
Authors: Lin Wang , Yan-jun Zhao , Xing-Zhen Zhang
Published: November 2022

The RAM system is the reactor control rod drive mechanism power supply system, which is the only power supply designed for the control rod drive mechanism. The RAM (control rod Power System) system of the Nuclear Power Plant only has two synchronous generators (RAM001/002AP) operating in parallel. On November 14, 2019,when one unit has an excitation system failure, due to the unreasonable setting of the protection value of the two units, the over-current protection of the unfailed unit precedes the loss of field protection action of the field lost unit, resulting the two generators in the RAM system tripping one after the other, therefore causing CRDM losing all power supply, resulting in the control rod dropping off and finally causing the automatic shutdown of the reactor, bringing serious consequences to the reactor and power grid. Therefore, the setting of synchronous generator protection in the RAM system is very important for the safe and stable operation of nuclear power plant. This report carries out the digital modeling and simulation of the RAM system of the Nuclear Power Plants, simulates various faults of RAM system synchronous generator RAM001/002AP and provides the important basis for verifying the settings of over-current protection and loss of excitation protection.
Pilot Application of RISMC Methodology on Specific SBLOCA Sequences of Passive NPP
Authors: Du Yun , Zhang Qinfang
Published: November 2022

In this paper, a typical passive nuclear power plant small break LOCA is taken as the research object. Based on the theory of Risk Informed Safety Margin Characterization methodology, the development process of the accident is studied. The specific sequence is selected to be studied. The important systems are discretely branched (such as Automatic Depressurization System), and the important uncertainty parameters are sampled (such as ADS valve and IRWST valve resistance coefficient). The event tree is re-established, and the success criterion of the head is modified. The fault tree is reestablished. After that the uncertainty parameters are sampled for specific sequences and the accident simulation of the whole plant is carried out for each condition. Thus, the frequency of each sequence and the conditional failure probability under the specific condition are obtained, and the core damage frequency (CDF) value based on RISMC method is finally calculated.

This analysis is mainly aimed at the configuration and sensitivity of ADS. The selected sequence 45,53,64 are generated from the risk significant core damage sequence S in the original model. Through the analysis and calculation of RISMC method, it is concluded that sequence 45 is ok sequence, and the assumption in the original model is too conservative. OK cases with epidemic proportions existed in sequence 53; sequence 64 is consistent with the original hypothesis, which is a CD sequence. Therefore, the calculated value of CDF is reduced to a certain extent.

This paper verifies the feasibility of RISMC method in application on passive NPPs and also proves that the method can remove unnecessary conservatism. more realistic assessment of the risk helps us quantify the safety margin of nuclear power plants better.
Numerical Study of the Effect of Different Tank Shapes on Liquid Sloshing Characteristics
Authors: Zhu Yuxuan , Lu Daogang , Li Donghao , Liu Yu
Published: November 2022

In nuclear power engineering, there are many tanks with irregular shapes (e.g., the cooling water tank of the containment, etc.), whose shapes are quite different from the traditional rectangular tanks. Under seismic excitation, the sloshing of the liquid in the tank produces the fluid-structure interaction (FSI) with the structure. It is crucial to study the sloshing characteristics of the water tank during the design of the reactor structure. There is a difference between the sloshing characteristics of irregularly shaped tanks and rectangular tanks, but there are few studies on this difference and the significance of the difference needs to be further investigated. To explore the aforementioned issues, based on the fluid-structure interaction theory and Housner’s equivalent mechanical model, the finite element models of the irregularly shaped water tank and the regular water tank (rectangular) are established by ANSYS software. The effect of the tank shape on the sloshing frequency and added mass of water in the tank is studied by modal analysis and calculation of the added mass. Further, the significance of the effect of tank shape on the natural and dynamic characteristics of tanks is investigated by comparing the variability of the sloshing frequency and the added mass. The results can guide the fluid-structure interaction problem in nuclear power engineering and provide a reference for the structural design of similar irregular water tanks, thus improving the safety and economy of reactor design.