bluuebunny/crossref_metadata_2025 huggingface dataset:
Design of Ecological Interface for Nuclear Power Plant Based on Cognitive Work Analysis
Authors: Yan Qike , Cheng Feng , Wu Hui , He Junzhou
Published: November 2022

Nuclear power plant accidents are an important public safety issue, and most accidents occur mainly due to unexpected events of the operator. However, traditional nuclear plant interfaces do not provide operators with the ability to resolve unexpected events. As an interface design method, ecological interface design can improve the situational awareness of operators and is widely used in nuclear power plants and other fields. At the same time, in order to improve the adaptability of the interface structure change, this paper introduces a formative method — Cognitive Work Analysis (CWA). This approach aims to describe the given constraints acting on a complex system from different perspectives and structural levels in order to better show how the system operates, and traditionally consists of five stages: Work Domain Analysis (WDA, which create a set of models that describe how complex systems work and then use those models to guide system design), Control Task Analysis (ConTA, which Build models for known repetitive tasks), Strategy Analysis (StrA, which observe known repetitive activity in more detail), Social Organization & Cooperation Analysis (SOCA, which designed to address team communication and cooperation constraints and boundary conditions), and Worker Competencies Analysis (WCA, which describe the level of cognitive control required by the operator to accomplish different system functional goals and tasks). This article will discuss the framework for the use of these five stages at the ecological interface.
Research on Optimization Method of Risk-Informed In-Service Inspection Strategy of RCP-RRA Connection Inlet Valve in Nuclear Power Plant
Authors: Li-Hui Chen , Zhang Yong , Shi-Jun Chen
Published: November 2022

This paper discusses the problems and risks exiting in the in-service inspection of RCP-RRA and connection inlet valve of nuclear power plant, and puts forward the method of optimizing the in-service inspection strategy according to the essence of risk-informed technology. Taking a domestic PWR nuclear power plant as an example, the application process of the optimization method is demonstrated through the combination of determinism and probability theory. The results show that it is acceptable to extend the disassembly inspection cycle of RCP-RRA connection inlet valve,verify the feasibility of the optimization method, and point out the impact of valve monitoring measures on the optimization method, so as to support the comprehensive decision-making and safety review of nuclear power plants.
An Insulation Structure With Noise Reduction Function
Authors: Yicheng Zhang , Pengfei Bian , Kai Zhang
Published: November 2022

The sound insulation performance of nuclear power related equipment and pipelines is of great practical significance for the safe operation of nuclear power equipment. At the same time, the surface of the equipment is often covered with an insulating layer to maintain the temperature of the nuclear power equipment. If the thermal insulation material can play the role of vibration isolation and noise reduction, it will undoubtedly have a positive effect on the vibration and noise control of nuclear power equipment. Aiming at the insulation layer structure using glass wool as insulation material, this paper studies the acoustic performance of the pipeline after installing the insulation layer by means of experiments and theoretical calculations. This paper verifies that the thermal insulation layer structure can effectively improve the insertion loss and sound transmission loss of the structure while meeting the requirements of the thermal insulation layer. At the same time, this paper conducts a parametric analysis study to study the effect of different structural variables on the sound insulation performance of the thermal insulation layer. The results show that the increase in the bulk density of glass wool, the increase in the thickness of the metal coat, and the additional damping layer all increase the insertion loss of the structure.
CFD transient simulation of natural convection characteristics based on detailed core rod bundles model in PLANDTL-DHX experimental device
Authors: Zhang Xueyuan , Zhang Yuhao , Guo Jingsong , Lu Daogang
Published: November 2022

If the plant blackout accident occurs in the pool-type sodium-cooled fast reactor, the decay heat of the core is discharged through natural circulation. The Plant Dynamics Test Loop (PLANDTL-DHX), an experimental device built in Japan, can simulate core coolant flow process and decay heat transfer phenomenon under decay heat discharged accident condition. In the present work, the numerical modeling of the experimental device is carried out based on the method of modular modeling and integrated coupling calculation, and the CFD commercial software FLUNET was used for calculation. The rod bundles of different forms in the core are modeled in fine detail. The initial conditions of transient are obtained under the steady boundary condition operation. Then, the change of key thermal parameters such as the temperature of the core and the temperature of IHX inlet and outlet are obtained by simulating the transient accident condition. In addition, there are obvious inner-flow and interflow in the core, meanwhile, the local backflow occurs at the core outlet. The influence of these phenomena on the heat transfer of the whole model is analyzed. The key results of simulation are compared with experimental data. The results can provide numerical references for the discharge of decay heat in the sodium-cooled fast reactor under power blackout accident.
Application of Risk-Informed Technology in Emergency Operating Procedure Development
Authors: Liu Liu , Liu Zhiyun , Yang Zongqiang
Published: November 2022

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Calculation of Detector Response Rate Based on BEAVRS Benchmark Using PWR-Core Analysis Code CORAL
Authors: Yang Wen , Liu Baolin , Yao Shiwei , Chao Fei , Li Xing , Qiu Jinrong
Published: November 2022

Three dimensional PWR-core analysis code CORAL developed by Wuhan Second Ship Design and Research Institute, which provides all functions required by PWR-core analysis calculation. These functions are neutron diffusion within the core and reflector, macroscopic depletion or microscopic depletion calculation analysis, multi-channel or sub-channel thermal-hydraulic analysis, one-dimensional heat transfer from nuclear fuel to the coolant, critical search by boron concentration or control rod position, integral and differential worth of neutron absorbers, neutron kinetics parameters for transient analysis, in-core neutron detector response simulation etc. CORAL is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to verify the computational capabilities of the reactor core analysis code, the BEAVRS benchmark is adopted as the research object. The BEAVRS problem is a benchmark problem based on real commercial PWRs with detailed material description, geometric information, core operating history, and detector measurement data. In this paper, the CORAL code is used to carry out physical analysis and calculation for the BEAVRS benchmark, and the response rate distribution of radial and axial detectors can be obtained. By comparing with the measurement results provided by the benchmark, it can be found that the calculation results of the CORAL code are in good agreement. This shows that the CORAL code can well simulate the detector response distribution in the core.
Experimental Investigation on the Characteristics of Bubble Departure Frequency of Subcooled Flow Boiling in Narrow Rectangular Channel
Authors: Chen Yong , Liu Hanzhou , Chen Deqi , Liu Haidong
Published: November 2022

Subcooled flow boiling in narrow rectangular channel is a very effective heat transfer method and is very important to nuclear reactor design and thermal hydraulic safety. The departure frequency of bubbles in subcooled boiling flow has direct influence on heat transfer. Subcooled flow boiling experiment is carried out to investigate the effect of different working conditions on the bubble departure frequency in the narrow rectangular channel. Because the test section is composed of quartz glass and Sapphire crystal Glass, it can be observed from four different directions to obtain high quality images. The experiment is conducted with mass fluxes of 400 to 1500 kg/(m2s) and heat flow density between 50∼300 (KW/m2). Besides, the system pressure is 100 kPa to 700kpa and inlet subcooled is 10 to 60 °C. Two types of nucleation sites can be detected from the experimental results: sliding growth and condensation. The bubble departure frequency of sliding is significantly higher than that of condensation. Besides, the effect of different working conditions on the bubble departure frequency in narrow rectangular channels was investigated. Finally, the departure frequency is analyzed and compared with previous models. The results show that the model proposed by Cole is in good agreement with the experimental results.
Verification and Validation of JMCT Monte Carlo Code in CPR1000
Authors: Ye Yaoxin , Zhao Jun , Bao Pengfei
Published: November 2022

The Monte Carlo method is essential in field of nuclear engineering and has been widely applied in reactor particle simulation owing to the rapid development of computer technology. Academic institutes and organizations of different countries have developed their own Monte Carlo code. The JMCT code is a Monte Carlo particle transport code developed by the China Academy of Engineering Physics. Before the JMCT Monte Carlo code is applied to the design of new nuclear power plant, the verification and validation (V&V) of the code on the operating reactors is necessary. In this work, the V&V of the JMCT code on criticality calculation and fixed-source calculation was performed on the operating CPR1000 reactors. First, the full core pin-by-pin model of CPR1000 was established and the criticality calculation was performed. Then, the 1/4 core model with irradiation surveillance capsules (ISC) of CPR1000 was established and the fixed-source calculation was performed. The calculation results of JMCT were compared to the reference code and the measured values in the ISC.
Comparative Model of Nuclear Power Plant General Layout Scheme
Authors: Huang Chong Guang , Wu Yan
Published: November 2022

Nuclear power plant sites are scarce resources for nuclear energy development. It is necessary to optimize the general layout scheme of nuclear power plant to reduce the project cost and operation cost. In the feasibility study stage of the project, the determination of the general layout scheme of nuclear power plant can provide design input for the land acquisition negotiation, geological prospecting, environmental assessment, site safety analysis and other topics. Then the rationality of the nuclear power plant project is proved more strongly.

The work content of general drawing runs through the whole process of power plant engineering construction, involving many aspects such as seismic geology, hydrometeorology, and bulky transportation. Among them, the comparison and selection of the general layout scheme of nuclear power plant is a key step of the demonstration in the early stage of the project, and it needs to be iterated with several topics in the early stage to be reasonably determined. Therefore, it is very important to find a method to compare and choose the optimal general layout scheme of power plant.

In this paper, a fuzzy comprehensive evaluation method based on AHP is proposed to establish an evaluation index system for the general layout of nuclear power plants. By comprehensive analysis of the main influencing factors of the project such as the self-human condition, three overall layout schemes are quantitatively evaluated from each dimension, and the optimal scheme is obtained.
The Research on Common Cause Failure Analysis and Countermeasures for Nuclear Safety Ventilation and Air Conditioning System
Authors: Peng Yongsen , Liu Zhengjie
Published: November 2022

By studying relevant regulations, standards and good practices, this paper discusses the typical types of common cause failures (CCFs) and general countermeasures in nuclear power plants, and summarizes the methodology, process and defenses for CCFs analysis of safety ventilation and air-conditioning (VAC) system in nuclear power plants. The typical CCFs are divided into four categories, namely functional dependency, spatial dependencies, inherent dependencies and human-related dependencies, and the corresponding defences are functional isolation, physical isolation, and diversity respectively. Comprehensive CCFs risk analysis of VAC system involves plant fault study, equipment FMECA analysis, PSA analysis, hazards analysis and other fields. For the identified CCF points, The optimal solution can be determined by combining factors which include technology maturity, feasibility, engineering cost, etc., and methods such as equipment diversity, physical isolation and functional isolation can be used to reduce or eliminate CCF points, to achieve as low as reasonably practicable (ALARP) risk of CCFs. The CCFs analysis methodology, process and defences summarized in this paper can provide reference for the CCFs of other nuclear safety supporting and auxiliary systems in nuclear power plants.
Analysis on the Determination of Coastal Nuclear Power Plant Platform Level
Authors: Huang Chong Guang , Wu Yan
Published: November 2022

In the project feasibility study stage, the determination of nuclear power plant factory dikes can effectively promote the rapid progress of the project, which can accurately calculate the quantity of construction conditions, the concrete position of the artificial slope, and other topics about factory ping elevation design input, effectively promote the progress of each project more forcefully demonstrates the rationality of the nuclear power plant engineering plan. It is very important to determine the plant floor elevation efficiently and reasonably. Based on this significance, a comparison and selection model for determining the plant floor elevation of nuclear power plant is established.

In the project feasibility study stage, the reasonably determined floor elevation will be used as the design input of other related demonstration topics, and there is a process of mutual iterative demonstration. The floor elevation of nuclear power plant needs to be determined based on various conditions of the site, so as to achieve the purpose of reducing the engineering cost and operation cost of nuclear power plant. This paper introduces the process of determining the floor elevation of a coastal nuclear power plant, and determines the floor elevation of a certain coastal nuclear power plant through comparison and selection analysis model. This analysis model is universal and important, and can provide the demonstration method and basis for determining the floor elevation of other coastal nuclear power plant sites. The analysis model mainly involves such factors as the amount of earthwork, the operating cost of circulating cooling water, and the slope of large transportation road.
Case Study on Significance Determination Process for NPPs Abnormal Events at Power
Authors: Chen Shijun , Zou Xiang , Chen Lihui , Luo Wenbo
Published: November 2022

With the research and application of Significance Determination Process (SDP) in China, it is urgent to carry out the detailed analysis of the significance of abnormal events at power, especially the NRC reactor regulation concept is being accepted to regulate the nuclear power plant in china. Based on Significance Determination Process (SDP) and the actual case analysis, this paper analyzes the significance of license operation event in a nuclear power plant with the current developed PSA model. The analysis process includes the initial significance determination, detailed analysis and external events/LERF analysis, finally determine the significance of license operation event, and provide a reference for the subsequent analysis of abnormal events.
Critical Heat Flux Experiments for IVR-ERVC Strategy Under the Pool Boiling Condition
Authors: Wang Gang , Kuang Bo , He Yihai , Liu Pengfei
Published: November 2022

Heat transfer limit is one of the main concerns of IVR-ERVC strategy. When the liquid level in the system is so low that natural circulation cannot be formed, the coolant near the outer surface of the reactor pressure vessel lower head is in the pool boiling state. In this research a one-dimensional full-height experimental facility was established to research the heat transfer limit (CHF) of ERVC under the pool boiling condition with a one-dimensional heating block, which is used to simulate the lower head of reactor pressure vessel. The experiment was carried out at different liquid levels and the results are compared with those of natural circulation experiment at the same liquid level. Experimental results show that CHF increases with the increase of the inclination angle of heating block. Meanwhile, the increase of liquid level is beneficial to the improvement of CHF. In addition, it can be found that the influence of flow path size on CHF is complex, and CHF does not change monotonically with the increase of flow path size. On the other hand, compared with the results of natural circulation at the same liquid level, the CHF values under pool boiling are relatively low. These results are expected to improve the understanding of IVR-ERVC strategy.
Review of Flow Characteristics in Narrow Rectangular Channels Under Marine Conditions
Authors: Wang Enpei , Li Lei , Wei Mingjun , Wen Yongsheng , Zhou Bowen
Published: November 2022

With the extensive application of offshore power plants and attaching importance of nuclear reactor system in marine industry, the narrow rectangular channels used for the compact heat exchanger in marine nuclear plant have drawn growing attention in recent years. Compared to some conventional pipes, the narrow rectangular channels boost the heat removal capacity efficiently and present the characteristics of compact layout, which have extensive application in compact heat exchangers. A significant number of experiments have been conducted on this topic. In this paper, a comprehensive historical review for narrow rectangular thermal-hydraulic characteristics under marine conditions is proposed, including an expansive number of experimental and numerical studies in open literatures for recent decade. The pressure drop and heat transfer characteristics are chiefly clarified in this paper. Additionally, the closely related studies of authors published in the last years are summarized in detailed. The followings are the significant aspects of this paper for dominant contents: (1) on the basis of a compilation of numerous distinguishing criterions, the definitions of narrow rectangular channels and marine motions are presented. (2) the flow characteristics of pressure drop and heat transfer characteristics in narrow rectangular channels under marine conditions are reviewed and summarized, basing on the authors’ published research achievements.
Integration and Future Development of Technical Standards for PWR Nuclear Pressure Equipment
Authors: Wu Feifei , Dong Zhenbang , Huang Jun , Liu Shangyuan
Published: November 2022

The development of nuclear power safety and the global procurement of nuclear power equipment have put forward an urgent need for the integration of nuclear power standards. In this paper, the development history and technical status of the localization of pressure equipment in China’s PWR nuclear power plants is introduced. This paper expounds the technology innovation and engineering experience of China’s independently-developed third-generation PWR nuclear power based on “HPR 1000” and “CAP 1400”, and the main goals, principles and basic ideas of the integration of technical standards for nuclear pressure equipment by sorting out and analyzing the list of topics in the fields of materials, design, manufacturing, welding, non-destructive testing and other professional fields involved in the integration of China’s nuclear pressure equipment standards, comparing with the technical topics involved in the current work of the SDO Integration Committee, and further discussing the differences and connections. This paper discusses the future development trend of nuclear pressure equipment in pressurized water reactor nuclear power plants and the demand for standardization of key technologies, and proposes the main problems that should be paid attention to and solved in the next step in the integration of nuclear pressure equipment standards.
Study on Fatigue Properties of Hydrided CZ2 Alloy
Authors: Huang Jieming , Gao Changyuan , Chen Liutao , Wang Xu , Tan Jun
Published: November 2022

CZ2 alloy is a new Zr-Nb zirconium alloy independently developed by China Nuclear Power Research and Technology Institute (CNPRI), which has independent intellectual property rights. Zirconium will react with coolant to absorb hydrogen in reactor service, which will degrade the performance of zirconium alloy and lead to failure. Therefore, in many studies of zirconium alloy performance, hydriding is often used to simulate burnup. In this paper, CZ2 alloy with hydrogen content of about 400 ppm was obtained by gaseous hydrogen charging method. The sample was heated to 500°C with the furnace in hydrogen environment and kept for 1.5 h. Then the fatigue tests of hydrided/un-hydrided CZ2 alloy and un-hydrided Zr-4 alloy under asymmetric axial pull-pull loading at 343°C were carried out to study the effect of hydride on the fatigue properties of CZ2 alloy. The results show that with the decrease of stress level, the difference between the fatigue life of CZ2 alloy un-hydrided and that of hydrided becomes larger, and the fatigue life of CZ2 alloy un-hydrided is higher than that of hydrided. The fatigue properties of un-hydrided CZ2 alloys are better than Zr-4 alloys. The two-parameter model and three-parameter model are used to fit the fatigue data of CZ2 alloy. It is found that the three-parameter model can better describe the fatigue life of CZ2 alloy.
Research on the Public Radiation Cognition Based on Social Psychology
Authors: Li Guanghui , Gong Yu , Bai Gang , Qu Yunhuan , Li Xiaoding
Published: November 2022

Radiation has the characteristics of being silent, colorless, odorless, invisible, and intangible. The public is deeply fearful and worried about radiation. Radiation has become “to mention the wolf’s name is to see the name”. Carrying out radiation science popularization is an effective way to crack the public’s fear. Research on public cognition of radiation is the basis of science popularization. Social psychology is a branch of psychology that studies public behavior dominated by psychology. Using social psychology can effectively analyze the psychology of public radiation cognition and improve the pertinence and effectiveness of science popularization. This paper investigates the current situation of public radiation cognition, and analyzes the existing problems of radiation cognition from the perspective of social psychology. It mainly includes that the public’s first impression of radiation is wrong, the public’s cognition of radiation is one-sided, ignoring the public’s needs, underestimating the public’s sense of resistance to radiation, underestimating the influence of wrong information, etc. In view of the above problems, this paper puts forward the methods to strengthen the pertinence and effectiveness of radiation cognition from the aspects of enhancing the effect of persuasion, paying attention to public needs, early public participation in radiation activities and reducing the impact of wrong information, so as to provide reference for radiation science popularization.
Study on the High Temperature Melting Treatment of Nuclear Waste Glass Fiber
Authors: Liu ChunYu , Wen YanLi
Published: November 2022

In abroad, the treatment method adopted for the wastes above is mechanical compression and storage [1–2]. In domestic, the wastes management methods could be summarized as follows according to the surface dose of the waste substance: (1) Waste filter cores with a surface dose rate greater than 2 mSv/h are fixed in a 400L steel drum with cement in the cement solidified line of nuclear auxiliary plant (NX), and one waste filter core is installed in a 400L drum [3–5]. After reconditioning, the waste volume is 0.4m3; By storing decay, the surface dose rate of some of the high dose rate level waste filter cores decreased to below 2 mSv/h, and then they were dried, super-compacted and cemented, which greatly reduced the amount of waste generated. (2) The waste filter elements with the surface dose rate below 2 mSv/h are packed into 200L steel drums and sent to the waste treatment auxiliary workshop (QS) for drying, super compaction and cement fixation. About 3∼4 waste filter elements are packed into a 400L drum. After preparation, the average waste volume of each filter core is 0.1m3. However, the treatment time of these methods is very long, the volume of waste after treatment is still relatively large, and the storage and isolation time is long. To sum up, this paper innovatively adopted high-temperature melting method to conduct glass solidification treatment on the simulated contaminated glass fiber, and by this method it forms a stable glass body with stable chemical properties. Moreover, it reduces the waste volume and directly forms a solidified body which is more convenient for treatment and disposal.