bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research on Operating Strategy of Total Loss of Essential Service Water System With RHR Not Connected in PWR Unit
Authors: Zhang Shengtao , Yi Ke 
Published: November 2022

Abstract
Essential Service Water System (WES) is part of the nuclear power plant cooling system which provides the final heat sink for nuclear power plants. Therefore, WES must operate stably, safely and reliably for a long time. The total loss of WES accident is a design extended condition and will result in the loss of the final heat sink of the unit. The consequences of the accident are severe. In order to deal with the accident quickly and effectively and ensure the safety and economics of the power plant in accident condition, it’s necessary to formulate corresponding treatment strategy to deal with the transient.
This paper developed a strategy for dealing with the total loss of WES with Residual Heat Removal System (RHR) not connected condition in certain PWR nuclear power plant. The structure of the WES system and the types of failures that may occur are analyzed, and thus the symptoms of the faults are obtained and the entry conditions for the operating strategy are determined. The effect of faults on unit equipment and safety functions and the impact on nuclear steam supply system (NSSS) control are analyzed in this paper. Combined with the unit design, the system and equipment for controlling and mitigating related safety functions are analyzed, and the mitigation method and the fallback strategy of the fault are determined. Thereby a complete operating strategy of total loss of WES with RHR not connected is obtained. In addition, this paper analyzes and evaluates the operating strategy by simulating thermal hydraulic calculation for the first time. The results show that without staff intervention Component Cooling System (WCC) temperature reached 55°C limits after running a few minutes. Based on the intervention of the operating strategy proposed in this paper, WCC temperature reached the 55°C limits when the unit was operated at about 4 hours and 55 minutes. The result shows that and the strategy can effectively alleviate the failure and provide sufficient intervention time for the operator to bring the unit to a safe state.
Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Study on Basic Device Specifications
Authors: Morita Koji , Liu Wei , Arima Tatsumi , Arita Yuji , Sato Isamu , Matsuura Haruaki , Sekio Yoshihiro , Sagara Hiroshi , Kawashima Masatoshi 
Published: November 2022

Abstract
A new subassembly-type passive reactor shutdown device is proposed to expand the diversity and robustness of core disruptive accident prevention measures for sodium-cooled fast reactors (SFRs). The device contains pins with a fuel material that is in a solid state during normal operation but melts and fluidizes during an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident. By rapidly transferring the liquefied device fuel into the lower plenum region of the pins via gravitation alone, the device can passively provide a large negative reactivity to the core. In this study, the nuclear and thermal properties of the device subassembly with metallic fuel were evaluated to determine the basic device specifications for proper device operation during ULOF and UTOP accidents. The results of the transient analysis of the ULOF initiating phase in a 750-MWe-class MOX-fueled SFR core showed that a conventional homogeneous core can maintain stable cooling of the core prior to coolant boiling in the driver fuel subassemblies. On the other hand, the negative reactivity required to terminate the event by device operation was found to be slightly larger in the low sodium void reactivity core than in the conventional homogeneous core.
Application of Vacuum Hot Helium Leak Examination Technology for ITER Blanket Shield Blocks
Authors: Li Weiping , Jiang Guoyan , Luo Wei , Fang Renjun 
Published: November 2022

Abstract
The Blanket Shield Blocks is an important part of the International Thermonuclear Experimental Reactor (ITER). In order to ensure its sealing performance, each shield blocks needs to be taken a vacuum hot Helium leak testing. A method for high-sensitivity examination using a vacuum thermal helium leak test system is proposed in this paper. The shield blocks can remove its own impurity gas and increase the activity of Helium atoms by increasing the environment temperature. The examination sensitivity is improved by increasing the pressure difference between the inside and outside of the shield block. The sensitivity of this method can realize the helium leak test with the system sensitivity of 5 × 10−11 Pa·m3/s under the conditions of examination environment temperature of 250°C, ambient pressure of 1 × 10−5Pa, and products internal helium pressure of 4MPa.
The results indicate that the technology has a high degree of automation and data traceability, which can avoid the influence of the product’s own outgassing on the examination, and has a high sensitivity to the examination of subtle air leakage.
Review on Electromagnetic Compatibility Research and Practice in Nuclear Power Plants
Authors: Yao Wenqing , Shang Xuelian , Chen Rigang 
Published: November 2022

Abstract
This paper studies the electromagnetic compatibility (EMC) for the instrumentation & control system, communication devices and electric equipments in commercial nuclear power plants (NPPs). It reviews the EMC research activities and practices for NPPs in United States, France, Russia and China. The paper first familiars the readers with electromagnetic interference (EMI) knowledge and examples that are most related to NPP EMC issues, including EMI source, coupling path and sensitive equipment. It then summarizes the historical research activities, focusing on the development of EMC standards for NPPs in those four countries. It is concluded that the current EMC standard systems are not uniform and lack of specific test procedures and quantification methods. They are insufficient for the nuclear industry to execute. At last, recent popular research topics for NPPs are reviewed including systematic EMC design, EMC design and control measures, EMC prediction and electromagnetic environment monitoring. Similar research work done in the general EMC community or by other industries has been introduced. It indicates that for NPPs EMC design especially at the system level is still immature, electromagnetic simulation can be a good tool to assist the EMC design and that EMC control regulations and environment monitoring become more and more necessary as wireless components and devices are being used.
Reliability Analysis of Off-Plant Power for Multi-Units
Authors: Honghao Chen , Cheng Ren , Jian Wu , Deyi Liu , Yang Luo , Shengjia Zou , Zilong Wang 
Published: November 2022

Abstract
Globally, multi-units are located at the same nuclear power plant site. After several serious nuclear accidents, the overall risk assessment of a nuclear power plant with multiple reactors has attracted much attention. Combined with the concept of PSA (probabilistic safety analysis) for multi-units, this paper analyzes the off-plant power supply configuration of 9 units in Qinshan Nuclear Power Base. In view of the 6KV emergency busbar directly affected by the instability of the off-site power grid, the correlation among several units is fully considered, and the exploratory research and analysis of off-site power grid reliability is carried out in Qinshan Nuclear Power Base. Taking the reliability of external power grid of Qinshan Nuclear Power Base as an example, the reliability of emergency bus and the influence range and contribution of various analysis factors to the evaluation results are quantitatively analyzed through fault tree modeling. The results of model analysis are in line with the actual situation. The analysis concept based on the multi-unit PSA methodology should be used in a wider range. For example, thematic analysis should be carried out for other highly correlated originating events (common cold source, equipment common cause failure, etc.), and the development and progress of multi-unit PSA should be promoted through practice with a point-to-surface approach. In the follow-up research, we should improve the model system, take the actual situation into full consideration, and develop a set of overall risk assessment system that can reflect the real-time risk level of multiple unit site.
Numerical Investigations of the LBE Flow and Heat Transfer Characteristics in the Helical-Coiled Tube Bundle
Authors: Shen Cong , Liu Limin , Liu Maolong , Gu Hanyang 
Published: November 2022

Abstract
Due to the advantages of compactness and enhanced heat transfer efficiency, the Helical-Coiled Once-Through Steam Generator (HCOTSG) is one of the best choices to be applied in the Lead-bismuth Fast Reactor (LFR). In order to investigate the flow and heat transfer characteristics in the shell side of lead-bismuth eutectic (LBE) HCOTSG for the purpose of safety analysis and optimization design, a CFD method for calculating the heat transfer and flow resistance in helical-coiled tube bundles is proposed. Based on available relevant experimental studies, the numerical model is validated. Validation results indicate that the liquid LBE cross flow through helical-coiled tube bundles can be well simulated using SST k-ω model and turbulent Prandtl number model. The maximum deviation for both friction factor and Nusselt number is less than 15%. After validating the reliability of the numerical model, geometry models with different helix angles of the shell side are established. Effects of helix angle on the flow resistance and heat transfer performances are presented. The flow and heat transfer characteristics in the helical-coiled tube bundle are numerically simulated and analyzed. This work provides a new method for studying the flow and heat transfer characteristics and optimization design of liquid metal H-OTSG.
Research on Public Communication of Small Reactor
Authors: Zhu Rongxu , Zhao Feng , Zhang Xiaofeng , Guo Jiandi , Zhang Meng 
Published: November 2022

Abstract
Based on the conclusions of small reactor environmental impact assessment and social stability risk assessment of small reactors, this paper conducts a special study on the public communication of small reactors. This article discusses the work content, scope and form of public communication of small reactor, and give suggestions on the scope and form of public publicity, public participation and information disclosure. It is matched with small reactor public communication theoretical system and operation manual.
The research on the public communication work of small reactor can provide necessary theoretical basis and technical support for the specific public communication work of small reactor. Provide decision support for local policies and construction departments to implement public communication. It is hoped that this study can provide some reference for the public communication of small reactors in the exploratory stage.
Applicability Study on Human Reliability Analysis for Severe Accidents in Nuclear Power Plants
Authors: Zhao Ming , Chen Shuai , Xu Jie , Zhou Jianzhang , Qing Tao 
Published: November 2022

Abstract
Three severe nuclear power accidents (Fukushima accident, Chernobyl accident and Three Mile Island accident) have proved that human errors in the mitigation of severe accidents will exacerbate the deterioration of accidents and cause immeasurable consequences. Therefore, there is an urgent need for finding human reliability analysis methods to effectively analyze and predict possible human errors in severe accident mitigation.In order to compare the applicability of the Human Reliability Analysis (HRA) method under the severe accident in the nuclear power plant, this paper put forward four criteria, qualitative analysis, quantitative analysis, traceability and availability, based on the characteristics of the personnel in the process of severe accident mitigation. Several commonly used HRA methods were compared and analyzed. The results show that the common HRA methods which applied to the analysis have shortcomings, so it is necessary to establish a new HRA method, or to improve the original methods to solve the issues under severe accident conditions and the IDAC model and Phoenix HRA methodology are considered better choice. Combined with computer programs,the IDAC model can simulate the process of accident and the dynamic response of personnel, which helps to explain the dynamic interaction characteristics of human behavior and environment, but the model still has some limitations in the aspects of uncertainty analysis and simulation data availability. While Phoenix method provides some Performance influencing factors and crew failure modes that can reflect the response characteristics of emergency personnel. This applicability study can provide guidance for the establishment of HRA method in severe accidents.
Study on Radial Flow Characteristics of U-Pattern 4-Wire and 7-Rod Bundles
Authors: Zhang Longzai 
Published: November 2022

Abstract
Wound wire bundles are often used in reactor structural components. The winding wire can not only prevent the vibration of the pipes, but also strengthen the heat exchange between the pipes. Because of its advantages in heat transfer, the wound wire bundle structure has been used in both core and OTSG.In this paper, the u-shaped 4-wound wire bundle structure is regarded as the research object, and four commonly turbulence models are used for calculation based on CFD program. Radial flow characteristics under different turbulence models are obtained through the given velocity inlet. What’s more, it is found that the SST k-ω model are shown better performance, when applied in the 4-wound wire bundle structure with the consideration of computer performance and calculation accuracy. The radial velocity of the three axial paths and the flow pressure at the center of the B-B slit show periodic distribution with a period of one quarter pitch. Due to the interaction between wire winding, there appears obvious flow pulsation in the radial direction of the central sub-channel in the structure of 4-wound wire bundle.
Study on a High Reliability Power Supply Scheme for Distributed Control System of Nuclear Power Plant
Authors: Hongpeng Tian , Lin Zhang 
Published: November 2022

Abstract
Distributed Control System (DCS) is regarded as the “brain” of nuclear power plant due to its function required in guarding nuclear safety. Power supply system is vital to supply electric power for DCS to maintain normal operation. Distribution design requirements of DCS power supply system are discussed firstly on analyzing the requirements of IEEE, IEC, etc. Combining with the comparative analysis of power supply schemes of DCS system in typical nuclear power plants based on CPR1000, AP1000 and EPR, key factors restricting the reliability of DCS power supply system are analyzed based on relevant design rules and standards in this paper. The method of improving the reliability of DCS power supply system is explored by analyzing the type of power supply, system design, equipment selection and design of electrical raceway. Thus a new high reliability power supply scheme for DCS system is presented based on mature equipment manufacturing technology. The scheme enhances the reliability, stability and flexibility of the distribution system by reducing the converting process on distribution path, increasing the design of cross-tie between divisions, and making full use of the advantages of DC distribution system. Through system operation analysis, equipment analysis and fault response analysis, this paper discussed the high reliability and feasibility of the new scheme, providing a new research direction for the design of DC system in nuclear power plant.
The Effect of Specimen Size and Surface Roughness on Tensile Strength of Nuclear Graphite
Authors: Guan Yuxin , Zhan Yingjie , Wang Shuiyong , Wang Qingwu , Ti Wenxin , Gong Bing 
Published: November 2022

Abstract
The tensile strength of nuclear graphite is an important parameter in evaluating and calculating the structural strength of High Temperature Reactor-Pebbled Modules (HTR-PM). The tensile strength of nuclear graphite with varied sizes and surface roughness was tested, and the size effect tensile strength formula was created. The results reveal that when the size of nuclear graphite increases, its tensile strength falls. However, in the 4 mm to 8 mm range, the size impact is less noticeable. The Weibull modulus m0.95 value is calculated based on the experimental data, and the size effect formula is established. The difference between the estimated value and the experimental value computed using the size effect formula is less than 10%. Tensile strength may be calculated using the size effect formula for nuclear graphite specimens of various sizes. The impact of surface roughness on tensile strength is linked to the grinding direction and specimen size. After a certain level of surface roughness, the tensile strength diminishes when the surface scratches are perpendicular to the tensile force. The stronger the effect, the smaller the specimen size. When surface scratches are parallel to the tensile tension, the tensile strength does not vary much as the surface roughness increases.
Experimental Study on Structural Vibration Characteristics of Single Rod in Different Flow States of Water
Authors: Yin Yibo , Wen Jiming , Yu Xinyang , Li Xin , Gao Puzhen , Tian Ruifeng , Sun Junshuai 
Published: November 2022

Abstract
In the nuclear reactor, flowing state of surrounding fluid will change structural vibration characteristics of fuel rods, which may lead to severe resonance and damage of fuel rods. Therefore, it is necessary to study structural vibration characteristics of fuel rods in different fluid flow states. In this paper, based on experiment, structural vibration characteristics of a single rod in different water flow states and different restraint modes are obtained by hammering method. The results show that different restraint modes have influence on dry mode, wet mode and structural vibration characteristics of the rod. Under condition of no excitation, when flow rate is lower than 1m/s, the response frequency and flow velocity change trend is the same under different restraint modes, response frequency of test rod with increases of flow rate. When flow rate is more than 1m/s, the response frequency varies with flow rate in different restraint modes. Under the fixed support, the excitation ways and excitation positions have no effect on natural frequency of rod. For simply supported, the excitation positions have great influence on the natural frequency of rod. It has reference significance for the safety design of fuel rods.
Evaluation System of Human-Factor Training Effect Based on Hierarchical Model
Authors: Zhong Xu , Xu Xia , Liao Xianglin , Tu Yichun , Song Yeshu , Liu Jiajin , Zhao Qian , Chen Xuefang 
Published: November 2022

Abstract
Nuclear power enterprises attach great importance to staff training and have invested a lot of resources to establish training and authorization systems based on Systematic Approach Training (SAT), including demand analysis, outline design, material preparation, training implementation and evaluation feedback five parts. However, the lack of effective training effect evaluation means in practice makes it difficult to continuously improve the training system. This paper takes human-factor training in nuclear power plants as an example, analyzes the characteristics of human factors training, such as complex training object, rich content and diverse methods, compares the Kirkpatrick model, Context, Input, Process, Product (CIPP) model and Context, Input, Reaction, Output (CIRO) model. The kirkpatrick four-layer model is selected as the theoretical basis to establish the evaluation system for the effect of nuclear power human factor training, including a closed-loop training evaluation process, five evaluation methods and corresponding tools. After a period of practice in a nuclear power operation team, this system has been proved to be able to achieve effective evaluation and continuous improvement of the human-factor training effect in nuclear power plants, and effectively improved the long-term performance of the operation team, the non human-factor mistake record has been more than 900 consecutive days. In view of the problems found in the system practice, the paper also puts forward suggestions for further improvement.
Probabilistic Analysis of Containment Overpressure Failure in Level 2 PSA
Authors: Liu Yu , Li Wenjing , Lu Wenkui , Wang Gaopeng 
Published: November 2022

Abstract
Containment is the final barrier to prevent any leakage and release of radioactive materials. The containment internal pressure may exceed the design value during severe accident, which could threaten the containment integrity. Therefore, the study on fragility and failure probability of containment under accident conditions has become a crucial issue in nuclear safety. In this paper, firstly, the PWR containment failure modes are identified based on relevant researches and engineering practices. Secondly, the containment failure caused by overpressure is divided into the structural failure and functional failure, which are discussed in detail. Finally, combined with both deterministic and probabilistic analysis methodology, fragility curve of PWR containment was evaluated as a step on the evaluation of containment failure probability due to the overpressure in Level 2 PSA.
Research on Accident Diagnosis Method of Reactor System Based on XGBoost Using Bayesian Optimization
Authors: Liu Yongchao , Li Xiangyu , Liang Biao , Wang Bo , Tan Sichao , Gao Puzhen 
Published: November 2022

Abstract
Traditional machine learning algorithms have problems such as overfitting, low accuracy, and difficulty in hyperparameter optimization when performing fault diagnosis.In order to improve the accident diagnosis ability of nuclear power plant reactor system, this paper combines Bayesian optimization (BO) algorithm with eXtreme Gradient Boosting (XGBoost) algorithm to develop a reactor accident diagnosis model.First, data preprocessing and feature quantity analysis are performed on accident data samples.Then, the BO algorithm is used to optimize the hyperparameters of the XGBoost model. Finally, the BO-XGBoost model is used to diagnose the operating conditions of seven nuclear power plants, and the diagnostic effects of various traditional machine learning classification algorithms are compared and analyzed.The results show that the BO-XGBoost model can achieve more efficient and accurate identification of reactor accident types, and the model has better generalization ability.This research can help nuclear power plant operators to accurately identify the types of reactor accidents, assist decision-making, and ensure the safe operation of nuclear power plants.