bluuebunny/crossref_metadata_2025_split huggingface dataset:
Application of Risk-Informed Technology in Emergency Operating Procedure Development
Authors: Liu Liu , Liu Zhiyun , Yang Zongqiang 
Published: November 2022

Abstract
Emergency operating procedures (EOPs) are used to instruct the operators to manage accidents in a nuclear power plant. Many plants are presently in the process of improving their EOP. It is important to recognize that the operators will have to deal with very unusual situations, use systems they are less familiar with, and possibly face unexpected plant behavior they have not experienced. In these situations the operators need reliable operating procedures to adequately respond to the complex and stressful situations and identify and take the appropriate actions. While typical operating procedures for many nuclear power plants are based on the deterministic analysis and the system design where important human actions and weakness are not being sufficiently identified. This report specifies the process by which plant operating procedures are developed and maintained using risk-informed technology so as to ensure that the operating procedures developed are technically accurate, comprehensive, explicit and user-friendly. Moreover, methods and benefits of risk-informed technology for developing procedures and verifying that the important human actions are addressed. Necessary enhancements and applications stressing the important human actions and weakness in developing operating procedures are also proposed which are of great significance in improving the safety operation of the plant.
Discussion on the Application of Risk Matrix Method in Regional Nuclear and Radiation Risk Assessment
Authors: Gong Yu , Zheng Jie-Ying , Xu Xing , Peng Hui , Qu Yun-Huan , Yang Li-Li 
Published: November 2022

Abstract
While nuclear technology benefits human beings, the ionizing radiation generated may also have adverse effects on human health and the environment. With the rapid development of social economy, nuclear energy and nuclear technology have been widely used, nuclear technology has become more and more close to people’s life, and nuclear and radiation safety has also attracted more and more attention from all walks of life. Therefore, while making effective use of nuclear technology, it is very necessary to do a good job of risk identification and safety management and control. Doing a good job in nuclear and radiation risk assessment is the premise of nuclear safety, nuclear security, and nuclear work. This paper is based on the risk matrix method proposed by the Australian and New Zealand standard “Risk Mangement, AS/NZS 4360:1999”, a qualitative risk assessment analysis method that can comprehensively assess the risk size of the possibility of hazard occurrence and the severity of injury. An attempt to carry out nuclear and radiation risk assessment in a city through the steps of risk source identification, risk possibility analysis, and risk consequence analysis. This paper briefly discusses the application of risk matrix method in regional nuclear and radiation risk assessment, and puts forward suggestions for the application of this method in regional nuclear and radiation risk assessment.
An Assessment Method of Digital Human-Machine Interaction Complexity in Nuclear Power Plants: EHMIC
Authors: Liu Xueyang , Cao Lu 
Published: November 2022

Abstract
A new way, entropy-based human-machine interaction complexity (EHMIC), was proposed and verified by real cases. Reasonable suggestions to reduce human-machine interaction complexity were given. Digital HMIC was studied using information entropy, and the difference between human-machine interaction and traditional system was controlled by digitalization. Task logic complexity, operation step complexity, information complexity, and knowledge complexity was determined as the main measurement indexes of digital HMIC, and they were integrated into a comprehensive index called HMIC. The index weight adjustment model was then established using the entropy evaluation method, and the HMIC calculation model was obtained through the Euclidean norm. This model was used to analyze HMIC during emergency operations of steam generator tube ruptures (SGTR) in a digital nuclear power plant. The quantitative analysis results were in accordance with the actual field situations. Results showed that the proposed EHMIC could provide good guidance to the selection of digital human-machine interaction programs, reasonable operator time distribution, regulations, and interface design and optimization.
Research on PSA Analysis for Equipment Degradation in Nuclear Power Plant
Authors: Yang Zhi , Zhang Sifan 
Published: November 2022

Abstract
It is nowadays a mainstream analysis method to use probability safety analysis (PSA) to support the nuclear power plants (NPPs) daily risk management and safety decision. In China, the NPPs are very experienced in doing PSA evaluation for different plant configurations such as equipment out of operation and failure, etc. In fact, however, the measures often used are not sufficient to reflect the real risk of the unit in such configurations, because there are numerous equipment degradation cases and few complete failure events in the NPPs. On this basis, the paper selects an event related to the main steam safety valve (MSSV) degradation in a domestic NPP as an example, collects failure data through experience feedback, and utilizes mathematical analysis and fitting function to determine the failure potential of the degraded equipment. By PSA calculation, the real unit risk under these configurations will be estimated finally. The purpose of this paper is to provide a new thinking and method for PSA analysis on degraded equipment in the NPPs.
Structural Analysis of a Reactor Vessel in a Sodium-Cooled Fast Reactor Under Extremely High Temperature Conditions
Authors: Yamano Hidemasa , Futagami Satoshi , Ando Masanori 
Published: November 2022

Abstract
To enhance resilience of next-generation nuclear structures, it is necessary to develop design methodology that mitigates impacts of failure caused by extremely high temperature conditions which might lead to a severe accident. The purpose of this study is to understand its deformation behavior under extremely high temperature conditions and to identify the areas that should be focused on to mitigate impacts of failure. For this purpose, this study has conducted a detailed structural analysis of a reactor vessel (RV) in a loop-type sodium-cooled fast reactor using a general-purpose finite element analysis code, FINAS/STAR. The RV was heated from the normal operation condition to the sodium boiling temperature in the upper sodium plenum for 20 hours, assuming depressurization. The analysis has revealed less significant stress and strain which were sufficiently lower than failure criteria. The upper body of the RV was identified as the important area in terms of mitigation of structural failure. The RV was eventually deformed downward about 16 cm, but it resulted in no failure. This effect contributes to maintaining the RV sodium level in a long term, thereby enhancing the RV resilience.
Practice and Discussion on Safety Categorisation and Classification of Defence in Depth Level 1/2 Functions Based on IAEA-SSG-30
Authors: He Fan , Ji Jiangwei , Xu Jiaoshen , Zhu Zengpei , Cai Kewei 
Published: November 2022

Abstract
The safety classification of structures, systems and components (SSC) in nuclear power plants (SSG-30) was issued in 2014 by the International Atomic Energy Agency (IAEA). Both the Hua-long Pressurized Reactor (HPR1000) and UK version of the Hua-long Pressurized Reactor (UK HPR1000) Safety Function Categorisation and System, Structure and Component (SSC) Classification methodology have been developed based on IAEA SSG-30 with consideration of different regulatory requirement. The SSG-30 is a high level guidance to perform the safety categorization and classification and there are still some issues that need to be analyzed in specific application such as failure consequence severity criteria and consequence analysis hypothesis. With the expansion of Defence in Depth (DiD) levels in nuclear power plant design, there are differences about function categorization requirements between different countries. This paper analyzes the differences of definition about DiD levels and the function categorisation requirements in different international standards. The IAEA SSG-30 methodology for application is evaluated to meet different regulatory requirements and the alignment or expansion is needed for application have been identified to determine the function categorisation in DID 1/2 levels. The principles of the HPR1000 and UK HPR1000 on DID 1/2 functions categorisation have been summarized and the subsequent introduce and discussion are about practices to apply these principles on special safety functions and safety features.
Effects of Bubbles Sizes on the Phase Distributions in a Bare Rod Bundle Channel
Authors: Han Sanjia , Qi Wei , Bu Shanshan , Li Zhenzhong , Chen Deqi 
Published: November 2022

Abstract
In order to clearly understand the phase distributions, the gas-liquid two-phase flow in a 5 × 5 bare rod bundle channel was numerically studied based on the Euler-Euler two fluid model with the CFX commercial software. In the two-phase bubbly flow simulation, the interfacial forces between gas-liquid phases including interfacial drag, lift, turbulent dispersion, wall lubrication, and virtual mass were considered. The experimental data was used to verify the numerical models. Different bubble diameters and MUSIG models in rod bundle channel were used in the numerical simulations, and the effects of bubble diameters on the phase distributions were carefully analyzed. It was found that the overall flow resistance increases with the larger bubble diameters. The bubble size significantly affects the void fraction distributions. The bubbles with the greater diameters tend to accumulate in the center of the rod bundle channel, while the bubbles with small diameters tend to appear in the rod bundle walls.
Research on Spacing Design Method of Condenser Half-Side Operation in Pressurized Water Reactor Nuclear Power Plant
Authors: Qingsen Zhao , Junqin Liang , Shuai Zhou , Jie Chen , Weipeng Zeng , Qiang Zhang 
Published: November 2022

Abstract
The condenser of Pressurized Water Reactor (PWR) nuclear power plant in China has many cooling tube leakage events, resulting in serious safety impact and economic loss. One of the main reasons is the fluidelastic instability caused by the large support plate spacing in half side operation. The corresponding calculation model is not given in the design specifications. Combined with the basic theory of cooling pipe vibration and comprehensively comparing different support plate spacing design and calculation methods, the mathematical model used in support plate spacing calculation in Heat Exchange Institute (HEI) standards is deduced. On this basis, the support plate spacing calculation model of condenser half side operation in PWR nuclear power plant is proposed. Based on a pressurized water reactor nuclear power project, the calculation and verification are carried out under four working conditions: Turbine Maximum Continuous Rating condition (TMCR), Winter Continuous Rating condition (WCR), Winter condenser half side operation condition (WHS) and Winter load shedding condition (WLS). In this project, the support plate spacing under winter condenser half side operation condition was 663.67mm, which is the most dangerous working condition under four working conditions. It is suggested that the final conservative support plate spacing should be determined according to the most stringent working conditions in the condenser support plate spacing design of nuclear power plant. Using the evaluation models of Alstom and Westhouse, the condenser support plate spacing of the PWR nuclear power plant project is checked and verified. The support plate spacing obtained by the mathematical model is relatively conservative and safe. The physical meaning of this method is clear and easy to use. It has certain guiding significance for the support plate spacing design of condensers in PWR nuclear power plants in China.
Safety Evaluation of Flowrate Dissymmetry on Nuclear Power Plant Operation
Authors: Li Dongyuan , Wang Jinkai , Xu Jiaoshen , Zhu Zengpei , Zhang Yaqi 
Published: November 2022

Abstract
It has been observed that the Reactor Coolant System (RCS) flowrate dissymmetry phenomenon (i.e. imbalance of primary flowrates between loops) appears in several Nuclear Power Plant (NPP) units. As an initial condition of fault studies, the RCS flowrate is very important to nuclear safety. The consequences of RCS flowrate dissymmetry are rarely investigated so far, and thus worthy of studying. This paper aims to analyze the safety margin of Pressurized Water Reactor (PWR) under specific accidents. A plant model is set up and accidents due to malfunction of a component in a single loop are simulated, in order to figure out whether Departure from Nucleate Boiling (DNB) is probable to happen. The results show that the Departure from Nucleate Boiling Ratio (DNBR) criterion is met for Partial Loss of Primary Flow accident with margins. For Main Steam Line Break accident, the flow dissymmetry has little impact on the accident consequences and the minimum DNBR meets the requirements. The Feedwater Line Break accident consequences are sensitive to the primary loop flowrate, while Departure from Nucleate Boiling (DNB) is not probable to happen with dissymmetry larger than 1500 m3/h. The results prove that the plant has enough safety margins and is able to operate under limited flow dissymmetry situation.
Crack Propagation Analysis of Dissimilar Metal Weld for Steam Generator Nozzle Safety End
Authors: Qi Shuang , Xiang Wenxin , Huang Ping , Ning Fangmao , Fan Minyu , Qi Lei , Chen Mingya , Xu Jun 
Published: November 2022

Abstract
According to the design load of steam generator, fatigue crack propagation of dissimilar metal weld joint were considered for the steam generator nozzle safety end of third generation pressurized water reactor (PWR) nuclear power unit. Based on finite element analysis, crack propagation analysis was carried out on dissimilar metal welds at the safety end of the circumferential plane crack joints on the inner surface of ellipse with a depth of 20 mm and a length of 120 mm. The mechanical response of the structures with defects was calculated in transient state of each design, and meanwhile the crack propagation was calculated during the whole lifetime. The fracture safety margin at the end of life of the component was verified. The results showed that, the maximum expansion of dissimilar metal weld was 2.2 mm in the direction of crack depth at the end of life, and then the fracture safety margin met the design specifications. In consequence, it provided theoretical basis and technical support for the acceptance of service defects of the third generation PWR nuclear power units, and even provided strong theoretical, methodological and data support for the structural integrity assessment of nuclear power plants, which had important theoretical significance and engineering application value.
Feasibility Analysis of Intelligent Instrument Application In Nuclear Power
Authors: Chen Shuzhi , Huang Meiliang , Shen Qimeng , Chen Jie , Liu Xiaoqiang , Wang Junli , Wang Weidong , Fu Yueming 
Published: November 2022

Abstract
To achieve the in-depth combination of informatization and intelligence has become a development trend in the nuclear power industry, and the digitization/intelligence of instrumentation is a big part. Based on the technical status of nuclear power instruments and industrial integrated circuits, this paper analyzes the feasibility and suggestions of intelligent/digital instrument application for nuclear power plant by combing the environmental conditions of nuclear power, intelligence/digitization of typical meters, testing and verification etc. basic work.
Analysis and Treatment of Trolley Overload Problem of Fuel Transfer System
Authors: Cheng Wei , Liu Zhengyang , Chai Panpan , Wang Zhiming , Wang Lei 
Published: November 2022

Abstract
Fuel transfer system is one of the key equipment which belongs to the fuel handling and storage system (PMC system) in PWR nuclear power plant. Its safe and stable operation is a prerequisite for fuel loading and unloading operation. During the first loading of a nuclear power plant in China, fuel transfer system had an overload problem of the conveyor trolley. The fuel loading operation was forced to stop on site, and the troubleshooting and treatment of this problem were carried out. This paper makes an in-depth analysis and research on this problem by using the Root Cause Analysis Method. After analyzing and troubleshooting the factors that may cause this problem one by one, we confirmed that the problem was caused by the over tolerance of the trolley straightness and the parallelism defect of some supporting rollers. Through adjusting the straightness of the conveyor trolley and the parallelism of the supporting rollers, this overload problem was finally solved successfully. All the results of the dry test, the 1m water test, the full water test and the fuel loading operation indicated that the conveyor trolley is in good working condition. The operating current of conveyor trolley is also stable within the design range.
Cause Analysis and Improvement of Drain Regulating Valve Blockage in Nuclear Power Plant
Authors: Kai Xue , Jing Li , Wensheng Hu , Chaofan Guo , Pengcheng Du 
Published: November 2022

Abstract
The secondary drain flow regulating valve of the Moisture Separator Reheater System of a nuclear power plant is frequently blocked, which will reduce the efficiency of the recovery and utilization of the drain water generated by the secondary reheater, thus affecting the efficiency of the thermal cycle. By comparing the drain control valves of other systems in the plant and the operation and maintenance of peer power plants, the power plant has determined the reasons for valve blockage: the main component of valve blockage material is Fe3O4, and the main reason for Fe3O4 is the precipitation of iron in the medium after saturation. Based on this, the temporary mitigation measures for on-line flushing are formulated. At the same time, the valve with blockage problem is redesigned and reconstructed, and the structure of the valve cage is improved. The design of the new structure valve cage is not conducive to the accumulation of impurities, and the cleaning of the valve cage is easier. The power plant will use the modified new valve cage in unit 2. After the transformation, the drainage capacity can meet the needs of the unit and the regulation performance is good. After the unit is at full power, it is observed for four months, and it is confirmed that the anti blocking ability meets the requirements.
Missile Risk Assessment for Nuclear Power Plant Under Super Typhoon
Authors: Feng Bingchen , Zhou Jie 
Published: November 2022

Abstract
Structures, systems and components (SSCs) of nuclear power plants are susceptible to wind-generated missiles under super typhoon, especially for water tanks and electrical equipment arranged in the open air. The tank of the conventional island demineralized water distribution (SER) system is the largest open-air tank of CPR1000 nuclear power plant, which has a high probability of being affected by missiles. To evaluate the missile risk for nuclear power plant under super typhoon, taking SER tank for example, an analysis using probabilistic safety assessment (PSA) method focused on typical accident sequences is carried out. Based on qualitative analysis of initiating events and mitigation functions required, the missile PSA model is established considered the failure probabilities of SER tank under different typhoon conditions. The quantitative evaluation is carried out from the aspect of accident mitigation and overall risk of the unit. The results show that the SER tank is one of the main weaknesses in accident mitigation and has a certain contribution to the overall risk of the unit. Therefore, the risk of missile for nuclear power plant under super typhoon should be concerned.
The Evaluation System Study on Layout Design of EMIT Activities in Nuclear Power Plant
Authors: Yikun Zhou , Nan Ding , Chengzhen Chen 
Published: November 2022

Abstract
The independently developed Generation 3 nuclear power technique of China, HPR1000, lacked an integrated and systematic design process facilitating Examination, Maintenance, Inspection and Testing (EMIT). Therefore, a set of corresponding design evaluation standards requires to be proposed. This paper discusses several evaluation dimensions with regard to spatial layout design for Examination, Maintenance, Inspection and Testing process, e.g., spatial accessibility for both personnel and equipment, operability for working personnel, spatial adequacy for all EMIT activities, convenience for staff to operate, monitor, surveillance, etc. These dimensions aim at proposing a relatively integrated and mature layout design evaluation system for Examination, Maintenance, Inspection and Testing. Quantitative standards are provided as reasonably practicable, according to regulations or codes of China, Europe or the United Kingdom. This paper does not discuss other EMIT aspects which are not related to layout design, e.g., isolability, radiation protection of specific working place during EMIT windows, equipment qualification, etc. Moreover, this paper also takes one of the typical components, Low Head Safety Injection pumps, of one PWR plant in China as a sample to introduce the application of this evaluation system. Within the introduction sample, the implication of each evaluation dimension proposed by this paper is interpreted during a specific and typical design scenario.
Research on Behavior of Migration for Molten Material in U-Al Alloy Fuels Under Severe Accident
Authors: Luo Yuejian , Deng Jian , Zhu Dahuan , Sun Hongping , Zhang Ming , Wu Xiaoli , Xu Youyou , Liu Lili , Hu Wenzhen , Xiang Qingan 
Published: November 2022

Abstract
U-Al alloy fuels have been applied in research reactors, and it is different from typical pressurized water reactors in geometry and components. This paper conducts a theoretical analysis on migration behavior of molten material in U-Al alloy fuels, and numerical simulation in SPERT research reactor, studying the migration behavior of molten material. The result shows that the number of cracks in cladding has significant influence on migration.