bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research on the Impact of Advanced Rule Design System on the Digitalization of Reactor Building Model
Authors: Wang Jin , Yang Hongyi 
Published: November 2022

Abstract
This paper proposes a more stringent method for customizing project rules. This method customizes the comprehensive management of the project and component reference database on the digital plant design platform based on general design codes, standards, item classification principles in nuclear engineering, digitalization requirements in reactor design, plant layout, project management, and material procurement and construction, etc. To improve the correlation between design specifications and digital power plants, suggest enhancing the data consistency among different design disciplines. Standardize the three-dimensional layout design of nuclear power plants, and ensure the consistency between the digital power plant model and the natural power plant. The rules are sorted out, analyzed, and transformed systematically in this paper. These rules include the naming and classification principles of items in nuclear power projects, model data composition structure, essential attribute content, component selection filters, material performance, model parameters, output content format, vital requirements in plant layout for reactor design, etc. This paper will finally form a systematic rule customization scheme through refinement and improvement. The parameters, such as process, operating conditions, materials, fluids, specifications, safety, quality assurance, seismic and radioactivity levels, as well as item naming rules, project database, component reference database, three-dimensional modeling, information integration, attribute inheritance, data extraction, and other regulations. This scheme can make the three-dimensional arrangement more standard, the operation steps more concise, and significantly reduce the designer’s attribute range of manual input. It can effectively promote the accurate and appropriate expression of the process and instrumentation design scheme in the reactor building model. Significantly shorten the project design cycle. Data integration and transmission between rules enable system attributes to be deeply inherited and automatic checking and judging of operating conditions parameters and pressure and temperature limits in physical properties of component materials. This scheme can make the three-dimensional layout more standard, the operation steps more concise, significantly reduce the attributes range of manual input by the designer, and obtain better application feedback in automatic drawing and material reports. These rules provide more comprehensive data support for coupling experiments, data integration, process simulation, and digital handover of different disciplines and depths.
Study on Numerical Simulation Method of Unsteady Deposition Characteristics of Particles in Surface Filtration
Authors: Ma Song , Sun Zhongning , Zhou Yanmin , Gu Haifeng , Li Lintai 
Published: November 2022

Abstract
As an important part of the ventilation system of nuclear facilities, the performance of filter is of great significance to the environment and personnel safety. With the operation of the filter, the filtration mode will change from depth filtration to surface filtration. In this paper, a numerical simulation method is developed for the unsteady deposition characteristics of particles in the surface filtration. The parcel injection model is applied to accelerate the formation and growth process of dust cake on the surface of fiber filter. A series of subroutines running in FLUENT environment are developed to enhance the capabilities of computational fluid dynamics code in ANSYS. The Brownian force is added to the DPM model to calculate the particle trajectory more accurately. The function that aerosol particles are intercepted and deposited on the surface of fiber filter to form dust cake is added. The porous media model is used to simulate the fiber filter and dust cake. The porosity and resistance coefficient change with the unsteady deposition of particles. Combined with the subroutine with FLUENT software, a numerical simulation method which can be used to predict the unsteady deposition characteristics of particles in the surface filtration is formed. The numerical simulation results are compared with the experimental results, and they are basically consistent. So the developed numerical simulation method can be used to predict the unsteady deposition characteristics of particles in the surface filtration, and can also provide a theoretical basis for the structural design of pleated filter.
Development and Application of a Support Device for Checking Leakage of Heat Exchange Tubes of Spring Base Supporting Condenser in the Form of Filling Water
Authors: Xie Jianning 
Published: November 2022

Abstract
The installation and removal workload of the original water filling support of the condenser is huge. For example, the water filling leak detection of the condenser during outage takes a lot of time on the main line of the conventional island, which cannot meet the requirements of the outage period. Moreover, the support can only be removed after the condensate side of the condenser is filled with water and emptied, which needs to be further optimized. Based on the stress analysis and load modeling calculation of the original water-filled support and its legs, the stress distribution and deformation of the water-filled leak detection support of the condenser are determined, and the design is improved according to the demonstration results. Through rapid fixed function of self-locking wedge device design, design of jack-up device and synchronization automated design of hydraulic equipment, a new type of spring water filling pedestal supporting type condenser leakage inspection support device was developed, and thus the condenser water filling leak quick installation and dismantling becomes convenient, safe and reliable, which significantly shortens the condenser leakage in the process of water filling to check support operation time. At the same time, the synchronous lifting of the water-filled leak detection support ensures the structural safety of the condenser during the process of condenser descent when the condensate side of the condenser is drained to the operating level.
Study on Passive Pulse Cooling Method of Secondary Side in PWR Nuclear Power Plant
Authors: Wu Zhenhua , Xu Junjun , Wang Weijin , Li Runcheng , Tang Qi 
Published: November 2022

Abstract
Currently, the supplementary strategy of PWR nuclear power unit to deal with the Station Black-Out (SBO) and Total Loss of Feed-water (TLFW) accidents mainly depends on mobile facilities and manual operation. Under the SBO together with TLFW (SBO+TLFW) beyond design basis accident, the self-sustaining time after the loss of active accident mitigation measures (safety injection / auxiliary feed-water / containment spray) of the nuclear power plant is limited, and the operation time left to the emergency team is relatively less. Especially in the case of external disasters such as super typhoon, the reachability of temporary facilities and personnel is challenged. By analyzing the available resources of the secondary side of the unit after SBO together with TLFW accident, this paper puts forward the passive pulse cooling method of the secondary side: the steam generators are divided into power SG(s) and cooling SG(s), the high-pressure steam of the power SG(s) is used as the passive power source, inject the high-temperature deaerated water stored in the deaerator into the released cooling SG(s), and the core is cooled by pulse water injection and intermittent exhaust. Through qualitative analysis of cooling mechanism, operation strategy and operation mode are given to put into services, and the simulation machine and CATHARE model are used for verification test and simulation calculation respectively. The results show that the two calculation trends are basically the same, which can delay core melting time more than 7 hours after SBO+TLFW accident. According to the probabilistic safety analysis (PSA) of a plant, the core damage frequency (CDF) is reduced by about 7%. The research results in this paper will be helpful to optimize TLFW response strategies and avoid putting into “Feed and bleed mode” too early and contribute to provide mitigation measures for delaying core melting under SBO+TLFW accident, and to gain time for putting external mitigation measures into service.
Modeling and Fault Diagnosis of Pressurizer Based on Bond Graph
Authors: Zhou Zhuoran , Peng Minjun , Wang Hang , Jiang Yingying 
Published: November 2022

Abstract
The bond graph can be used to simulate and model conservation processes in different energy domains. In this paper, based on bond graph, the pressurizer was modeled and diagnosed at full power. Two parameters of pressure and mass flow were selected. Firstly, the bond graph of pressurizer was established. The pressurizer was divided into three zones: the fluctuating zone, the saturated liquid zone and the gaseous zone. The conservation equations of mass and energy were given respectively for these three regions. Secondly, the conservation equations of mass and energy were combined with the conservation of volume, and the model was constructed according to the combination rules. Then, according to the established bond graph model, the time causality diagram was established to determine the influence relationship between the coefficients and listed the fault characteristic matrix. Meanwhile, the reasoning structure between the coefficient was realized by the expert system G2. Finally, the fault data simulated by simulator was imported into the bond graph model. The failure of the safety valve, the spray valve and the immersion type heater were introduced to verify the modeling and diagnosis effect. It concludes that the diagnosis effect is consistent with the imported fault. Compared with the data-driven method, the bond graph method is more reliable, and compared with the expert knowledge method, the bond graph method can diagnose unknown faults and has a wide range of applications. In addition, a new method for nuclear power plant simulation modeling was provided. This was a new attempt to model bond graphs outside the research of mechanic, electronic and hydraulic.
Selection of Complex Accident Sequences for Sodium-Cooled Fast Reactor Design Extension Conditions
Authors: Jiang Jingke , Zhao Nanfei , Yang Dandan 
Published: November 2022

Abstract
Design Extension Conditions (DEC) are accident conditions beyond design basis for nuclear power plants, including complex accident sequences (DEC-A) and severe accidents (DEC-B). As required by the nuclear safety regulation HAF102-2016, DEC must be considered by identifying additional accident scenarios and develop practical prevention and mitigation measures for such accidents in design process for nuclear power plants. The aim is to enhance nuclear power plant’s response to beyond design basis accidents or multi-fault accident. It’s to contribute to increase the design requirements and expand the design consideration for nuclear power plants to consider DEC in design process, which may increase the design difficulty and construction and operation costs for nuclear power plants. The selection of DEC is great significance to balance the safety and economy for nuclear power plants.
Compared with DEC-B, the DEC-A is a newer concept, and the mature cognition of DEC-A has not yet been reached. The selection method of DEC-A for sodium-cooled fast reactor is analyzed systematically, which includes how to define, identify and screen DEC-A. Taking a sodium-cooled fast reactor as an example, DEC-A are identified and screened in the method of the combination of probabilistic safety assessment, determinism and engineering experience, and the entire selection process of DEC-A is performed. Besides, It is analyzed and discussed about the application of DEC-A in safety analysis and safety design of sodium-cooled fast reactors. Existing study results will provide theoretical reference for determining DEC-A lists for similar reactors.
Experiment Study on the Effect of Nozzle Shape on Liquid Jet Breakup
Authors: Sun Guofu , Zhan Yi , Okawa Tomio , Aoyagi Mitsuhiro , Uchibori Akihiro , Okano Yasushi 
Published: November 2022

Abstract
When a liquid sodium leakage accident occurs in a sodium-cooled fast reactor, the injected sodium collides with structures to produce splashing droplets, which can result in a violent combustion. According to previous studies on circular nozzles, the amount of splash is affected by the state of the jet at the moment of impact. However, the outlet shape of damaged area is hardly to be circular; and meanwhile it influences the flow pattern of jet a lot. Considering about this, in the present work, high-speed cameras were used to observe the jet discharged from oval nozzles vertically downward to investigate the falling process of the jet. The result shows that surface wave appears on the jet and within a certain range of flow velocity it can be observed obviously, meanwhile accelerate the breakup of jet.
A Proof-of-Concept Application of Fast Fracture Evaluation for the Primary Loop Section in Nuclear Power Plant
Authors: Zheng Zihao , Qiu Wenxin , Dai Yang , Min Jiesheng 
Published: November 2022

Abstract
The primary loop in nuclear power plants (NPPs) consists of piping and pressure vessel, which operate under a high level of pressure and temperature. During the long-term service, these sections need to have their integrity analyzed to check whether they might be prone to degradation, especially when exposed to radiation. In-service codes and standards such as the RSE-M and the ASME Section VIII enclose rules for the analysis of defects in piping systems. Although geometries for pipes are well-known and the fracture analyses methods are all today quite mature, it still takes an amount of time for structural engineers to rebuild models of different defects on piping section and the evaluation results varies with different modelling methods.
Hence, this paper provides a proof-of-concept tool for fast fracture evaluation, specific to the brittle fracture induced by thermal fatigue and radiation in the primary loop. In this application, the procedure of finite element method (FEM) analysis and fracture-related criteria is well integrated, based on the open-source FEM code code_aster and multi-tool platform SALOME. It automates several steps of primary loop section simulation, including mesh generation, calculation file configuration, post-processing and reporting results.
Furthermore, the verification and validation tests are built to ensure the robustness and reliability of the software and productivity.
Maintenance Strategy Optimization Based on AP-913 in Relay Protection Field
Authors: Zhao Yu , Li Ping , Tian Kunpeng , Fei Xingwei 
Published: November 2022

Abstract
AP-913 (Equipment Reliability Process Description) describes an equipment reliability process offered to assist member utilities to maintain high levels of safe and reliable plant operation in an efficient manner. This document based on AP-913 process document, this paper generally describes its determination by logic and analysis method. Combining with the types of failure and consequences in the actual problems, it created a more simplified maintenance strategy optimization mind maps (including the system stability, the design necessity, the economic efficiency, the equipment reliability four factors, refining the equipment and benefit related factors, its relationship and influence between the four are clarified), that can raise efficiency of maintenance and enhance the stability of the system. Through its application in the maintenance cases, it takes some cases for example: Increase the PM project of replacing the power board card of the transformer group, shorten the PM project cycle of the transformer group power transmitter, and change the control operation mode of the main transformer cooler;cancel the design of outdated manual working frequency excitation equipment and its PM items, cancellation of electric disk cabinet timer, etc. It is proved that the reliability of the equipment is improved after the implement of the maintenance strategy optimization. Time costs and labor costs are saved, and the cost of equipment procurement is decreased too. Responding to control the company’s cost positively, some ideas at the same time also can provide reference for other similar plant maintenance work.
PSA Application for UK HPR1000 ALARP Demonstration
Authors: Wang Zhiwen , Yang Chunju , Zhang Bing , Wang Jinkai , Sang Huina , Xiao Lingmei , Wang Fei , Wang Ming 
Published: November 2022

Abstract
UK version of the Hua-long Pressurised Reactor (UK HPR1000), designed by China General Nuclear Power Corporation (CGN), has been successfully completed the Generic Design Assessment (GDA), with the issue of a Design Acceptance Confirmation (DAC) from Office for Nuclear Regulation (ONR) and a Statement of Design Acceptability (SoDA) from the Environment Agency on 7 February 2022. In the UK GDA process, one of the main fundamental requirements is that the nuclear safety risk should be reduced to ALARP (As Low As Reasonably Practicable). During the UK HPR1000 design process, Probabilistic Safety Assessment (PSA) has been widely used as a key tool to inform the design and evaluate risk levels, and to demonstrate that the nuclear safety risk has been reduced to ALARP.
The process of PSA ALARP demonstration is established in accordance with the methodology that reflects the UK context requirements and consistent with relevant good practice. The process is fully applied by using PSA to systematically review the risk insights and to support other topic areas. After the systematic review of PSA risk insights from internal events Level 1 PSA, internal fire Level 1 PSA, internal flooding Level 1 PSA, external hazards Level 1PSA, external flooding Level 1 PSA, seismic PSA, Spent fuel pool PSA and Level 2 PSA, the related conservatisms, potential design improvements have been identified and addressed.
Current UK HPR1000 PSA results and risk insights show that the risk level of UK HPR1000 design is relatively low and the risk of UK HPR1000 design has been reduced to ALARP in GDA stage.
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.