bluuebunny/crossref_metadata_2025 huggingface dataset:
Research on Simulation of Emergency Evacuation Outside Nuclear Power Plants Based on Genetic Algorithm and Variable Neighborhood Search Algorithm
Authors: Yan Jin , Zhang Jiemin , Cao Qun , Wang Mengxi , Long Liang , Xue Na
Published: November 2024

This paper focuses on the evacuation process of the public during off-site emergency, and studies a simulation method for off-site emergency evacuation based on genetic algorithm and variable neighborhood search algorithm. And this model is used to calculate and analyze the off-site emergency evacuation plan for a nuclear power plant, based on information such as population, vehicles, assembly points, settle places, transportation networks, and radiation dose fields around the nuclear power plant. The optimization objectives are evacuation time, collective dose, or individual maximum effective dose. The consumption time of the evacuation process and the dose exposure received by personnel during the evacuation process are provided. And compare the results with traditional evacuation simulation algorithm. The results indicate that emergency evacuation plans should not only consider the evacuation time, but also comprehensively consider the dose received by personnel during the evacuation process. In addition, this paper analyzed the impact of adding private cars on the evacuation process, and the results showed that when adding more than 50% of private cars to the number of buses, the impact on evacuation time, collective dose, and individual maximum effective dose significantly increased. In addition, when the evacuation time is taken as the optimization objective, the evacuation time of the optimized evacuation plan through genetic algorithm combined with variable neighborhood search algorithm is shorter than that of the traditional simulated evacuation plan. This algorithm can generate a more optimized evacuation plan.
Development of a Nonintrusive ROM for Parametric Study in the RPV Lower Plenum
Authors: Shang He , Peng Minjun , Xia Genglei , Zhang Shiqi , Zhu Shijie
Published: November 2024

System thermal hydraulic codes are extensively used to assess the response of large-scale systems due to their computational efficiency in simulation. Nevertheless, they cannot simulate inherently multi-dimensional physical phenomena because of the one-dimensional approach adopted. A common example is the multi-dimensional flow in the reactor pressure vessel lower plenum. One solution is to employ computational fluid dynamics codes to simulate the lower plenum and obtain the required multi-dimensional flow. However, computational fluid dynamics computation is typically time-consuming, introducing a new challenge. In this paper, a nonintrusive reduced order model for steady state calculation of two-dimensional lower plenum has been developed using proper orthogonal decomposition and sparse grids methods to overcome this challenge. The input of reduced order model is the coolant mass flow rate at the inlet of the lower plenum, and the output is the velocity field distribution in the lower plenum. The range of the inlet flow rate is from 500 kg/s to 600 kg/s, and all data in snapshots are generated through repetitive simulation under different sampling points using OpenFOAM. The performance of the constructed reduced order model is validated on a test set including 16 new sampling points. The results indicate that the established reduced order model can rapidly provide velocity field distributions in the lower plenum for any inlet flow rate ranging from 500 kg/s to 600 kg/s, and the errors are all acceptable to meet simulation requirement.
Design and Calculation of Nuclear Heating System Utilizing a Molten Salt Heat Storage Circuit for Waste Heat Absorption and Peak Regulation
Authors: Chong Haofang , Xue Ruojun , Wang Sen , Cao Yaowu , Jing Furu
Published: November 2024

The utilization of nuclear energy heating systems is prevalent in the heating sector. However, the peak regulating capacity of nuclear heating systems is typically inadequate. To improve it, a nuclear energy heating system integrated with molten salt energy storage is devised. Molten salt energy storage systems are widely adopted in thermal power units and photovoltaic power generation to regulate peak load and store excess energy. By integrating the molten salt energy storage system into the nuclear energy heating system, surplus heat from the reactor can be temporarily stored during periods of low demand in the central heating network and released when there is high demand, enhancing the self-regulating peak capacity and operational flexibility of the system. Furthermore, coupling with the molten salt energy storage system allows for partial replacement of certain functions of the original residual heat removal system by absorbing shutdown waste heat at the initial stage of reactor shutdowns and providing it to either the central heating network or other users. In this paper, we establish a self-programmed one-dimensional calculation model for evaluating whether using molten salt to absorb initial waste heat from reactor shutdowns in a nuclear heating system is feasible and if this absorbed heat can be applied to supplementing central heating networks; additionally, we conduct preliminary calculations on utilizing molten salt for peak regulation purposes within the central heating networks, to preliminary verify the usability of the design scheme.
Effect of Corrosion Behavior on the Heat Transfer Performance of Fuel Cladding in Lead-Cooled Fast Reactors
Authors: Feng Wenpei , Yang Shangdong , Yu Hongxing , Lu Xingqing , Zhang Xue , Gu Zhixing
Published: November 2024

In lead-cooled fast reactors, preventing corrosion of materials like stainless steel by liquid metal is crucial. The protective oxide film on stainless steel surfaces, formed to mitigate corrosion in lead/lead-bismuth systems, can, paradoxically, degrade heat transfer in cladding during high-oxygen concentration or normal operation. This study investigates the impact of corrosion on heat transfer, focusing on steady-state conditions. By analyzing temperature changes with and without corrosion, the study aims to understand the balance between oxide film formation and heat conduction. Additionally, it assesses corrosion effects under accident conditions to enhance safety in lead-cooled fast reactors. The findings contribute valuable insights for designing and operating cooling systems, ensuring the safety and reliability of nuclear energy systems.
A Nonlocal Meso-Scale Damage Model for Nuclear Graphite
Authors: Ma Yan , Qian Yue , Hao Yuchen , Lin Musen , Li Yue , Wang Jinhua , Wang Haitao
Published: November 2024

Due to its excellent nuclear physical properties and high-temperature resistance, graphite is widely used in many areas of the nuclear industry. In high-temperature gas-cooled reactors (HTGRs), graphite is used not only in the core structural components as the neutron moderator and reflector, but also as the matrix material for the fuel elements. During the operation of HTGRs, the core structure and fuel element are subject to various static and dynamic loading. Therefore, it is particularly important to study the reliability evaluation of nuclear graphite for the safe operation of reactors. As a typical brittle material, the quantity, size and shape of internal defects of nuclear graphite have strong randomness, which leads to a large discrete strength of nuclear graphite. It is more reasonable to use a probabilistic method to evaluate the strength of nuclear graphite. At present, the dominant approach to the probabilistic evaluation of nuclear graphite is based on finite element methods, such as the Burchell model. However, compared with the meshless method, the finite element method make it difficult to handle the stress singularity brought by the displacement discontinuity, the computation complexity brought by the crack bifurcation and the non-locality of failure. As an emerging meshless method, Peridynamics has been adopted by most scholars. In this paper, we develop a uniaxial tensile fracture model of nuclear graphite based on the ordinary state-based Peridynamic theory. After that, a random number and size of initial defects are added to the nuclear graphite according to the pore-size distribution in the graphite microstructurally parameters. In order to verify the reliability of the model, this paper compares the prediction results of the model with the experimental data on two typical graphites using the two-parameter Weibull function. Besides, this paper compares the model in this paper with the Burchell model, and the results show that the spherical defect shape used in this paper is more suitable for predicting isostatic graphite.
Bubble Features on SiC Surfaces in Flow Boiling
Authors: Bing Tan , Ren Haoliang , Sun Rulei , Cheng Songbai
Published: November 2024

The study of bubble dynamics is crucial for understanding the boiling heat transfer process, predicting heat transfer efficiency, and preventing film boiling issues. The research aimed at experimental exploration of bubble characteristics on silicon carbide (SiC) surfaces during flow boiling at subcooling levels of 4K to 12K and fluid flow rates between 0.15–0.3 m3/h, all conducted under atmospheric pressure. The investigative method harnessed a U-net artificial neural network to meticulously identify attributes of the bubbles, focusing on properties such as orientation, perimeter, effective diameter, roundness, and aspect ratio distribution. Key parameters for the working fluid were set within the specified ranges for subcooling and flow rates, ensuring the experimental conditions replicated flow boiling at atmospheric pressure. Utilizing U-net as a tool allowed for precise recognition of bubble features on SiC surfaces, which are the primary subject of this study. The outcome of the experiments, facilitated by the neural network, successfully mapped the distribution of various characteristics of bubbles on SiC surfaces under flow boiling conditions. This provided valuable experimental data and insights that aid in the further understanding of SiC surface behaviors in flow boiling processes.
Research on Sealing Performance and Electrical Performance of Electrical Penetration Assembly in Nuclear Reactor Containment
Authors: Wang Yanlu , Yu Xinli , Cao Xiaxin , Liu Yu
Published: November 2024

Electrical penetration assembly (EPA) are electrical conduits mounted on the containment that serve to maintain the integrity of the containment pressure boundary and provide a path for cables to traverse the containment. Currently, low-voltage electrical penetration assembly (LV-EPA) is mainly made of organic materials to realize the sealing function, and it is not clear whether the performance of organic sealing materials will be degraded in the event of a nuclear reactor under extended design conditions (DEC-B). In this paper, a representative original-size LV-EPA is selected as the research object, and the equipment life test and the test for measuring the insulation and sealing performance of the LV-EPA under severe accidents are carried out by constructing a test setup that simulates severe accident conditions. The results show that: under the condition of 200°C and 0.6MPa, with the increase of test time, the electrical performance of LV-EPA remains good, but the sealing performance will be degraded step by step; after 72h simulated DEC-B test, the sealing performance and electrical performance of the equipment satisfy the standard requirements.
Study on Application of the Single-Curvature Polyhedron Hydro-Bulging Technology in Metallic Reactor Containment
Authors: Dong Jianling , Huang Xingcheng , Chen Ke , Ning Bohao
Published: November 2024

The integral RPV configuration and layout nuclear reactors eliminate the loop piping and the externally located steam generators, pumps and pressurizer with their individual reactor pressure vessels. Hence, the footprint of the containment can be greatly reduced. This size reduction makes them possible to use small size and high pressure vessels as their containments. Compared with cylindrical pressure vessels, spherical vessels have the advantages of higher pressure bearing capacity and material saving. So for integral RPV configuration reactors, using spherical vessels as their containments and reactor building as a barrier to protect the containment from attack from aircraft is a promising configuration and layout plan.

The traditional spherical vessel technology adopts presses and dies to form the steel sheet blanks into double-curvature orange petals, then to assemble and weld the petals to obtain a spherical vessel, the process is complicated, long production period and higher cost.

The single-curvature polyhedron hydro-bulging technology to manufacture spherical vessel is a new technology for manufacture spherical vessels, which has been used to manufacture spherical vessels in many fields. But up to now, there is no case that the technology is used in the nuclear industry. This paper tries to explore the application of the single-curvature polyhedron hydro-bulging technology for manufacturing spherical vessel in manufacturing steel spherical containment vessel of the nuclear power plant.
Overview of the Application of Thermal Management in Industry and the Application Prospect in the DCS of the Nuclear Power Plant
Authors: Xu Youyou , Deng Jian , Qiu Zhifang , Zhu Dahuan , Wang Dongwei , Han Wenbin
Published: November 2024

Reactor digital control system (DCS) is the brain and central nervous system of nuclear power plant. It is the most complex system of nuclear power plant, which can ensure the safety and stable operation of the nuclear power plant. The DCS in nuclear power plants can be divided into safety-related DCS and non-safety-related DCS. With the improvement of digital, intelligent, integrated and miniaturized technology of the DCS in nuclear power plant, thermal management is becoming an increasingly integral part of advanced DCS system design. In electronic devices, temperature is one of the key factors affecting the performance of the electronic components. In the field of electronic heat dissipation, there is a famous ten-degree rule: the reliability of electronic devices is closely related to the temperature. Within a certain temperature range, the temperature of electronic components rises by 10 degrees, its reliability will decline by 50%. Therefore, effective thermal management of electronic equipment is particularly important. Usually, thermal management technology mainly focus on three key elements: heat source, heat sink and the channel to removing heat. Large amount of researches about these three elements were carried out in order to meet the requirements of application scenarios and various heat indexes. On one hand, thermal management can reduce the failure probability of the DCS, on the other hand, more efficient thermal management can reduce the weight, volume and energy consumption of the DCS. Firstly, the application of thermal management technology in various industrial fields is systematically reviewed and summarized in this paper. Then, combined with the three elements of thermal management mentioned above, the application scenario of thermal management in DCS of nuclear power plant is analyzed. Systematic thermal management research on DCS of nuclear power plant aims to lay the groundwork and support for the subsequent application of advanced thermal management technology in DCS of nuclear power plant, so as to reduce the weight, size and the energy consumption, reduce the space occupied by DCS and improve the reliability of the instrument control system.
Application of Artificial Neural Networks in the Optimization of Feedwater Heater Models
Authors: Dong Shubiao , Xiaoyu Zhang , Qianping Zhang
Published: November 2024

The traditional lumped parameter model for feedwater heaters is insufficient to accurately predict changes in heat transfer performance caused by factors such as aging and fouling during power plant operation. This study established an operational database consisting of 68 million datasets and employed it to train an artificial neural network model. This neural network model establishes a mapping relationship between operating time, thermal hydraulic conditions, and heat transfer efficiency correction coefficients. Subsequently, the neural network model was employed to optimize the heat transfer calculation process of the lumped parameter model, and a new composite model was formed after integration. Finally, the optimization effect of the model was tested using operational data. The results show that compared with the lumped parameter model, the composite model exhibits higher and more stable computational accuracy at different operation times. This means that the composite model can, to some extent, reflect the changes in heat transfer performance of the feedwater heater during operation.
Experimental Study on Two-Phase Flow Instability in Parallel Helically Coiled Tubes Under Rolling Motion
Authors: Wang Ruohao , Qi Chao , Ren Jiaxing , Qiao Shouxu , Tan Sichao , Tian Ruifeng
Published: November 2024

Parallel helically coiled tubes are extensively employed in heat exchangers of floating nuclear platform owing to their compact configuration and exceptional heat transfer efficiency. The helical structure of the coiled tube leads to the occurrence of secondary flow, making the flow instability phenomena more complex. Under ocean conditions, the helically coiled tube will suffer from more serious two-phase flow instability. Thus, research on the two-phase flow instability of parallel helically coiled tubes under rolling conditions is crucial.

The present study experimentally investigates the two-phase flow instability in parallel helically coiled tubes under static and rolling conditions. Two vertically parallel helically coiled tubes which have 8mm hydraulic diameter are used in the experiments. Ledinegg flow instability and Thermal Oscillation (TO) are observed under static conditions. The single-phase oscillations, Ledinegg flow instability and TO are observed under rolling conditions. The experimental phenomenon under static and rolling conditions are compared.

The stability map of parallel helically coiled tubes is established. The Fast Fourier Transform (FFT) method are used to investigate the evolution characteristic of the flow instability under rolling conditions. The variation of amplitude spectrums indicates that impact of rolling conditions on flow instability diminishes as heating power is increased. Finally, the parameter effects of rolling angle and rolling period on two-phase flow stability are indicated.
A Summary of Historical Evolution of Radiation Protection Policy From ICRP
Authors: Huang Yufei , Zhang Zongyuan , Liu Ling , Wang Yu , Guo Jingni , Wei Liqiang , Xie Feng , Tong Jiejuan
Published: November 2024

The International Commission on Radiological Protection (ICRP), as an unofficial organization dedicated to advancing the science of radiation protection, plays a crucial role in formulating radiation protection guidelines, providing scientific advice, and establishing principles in this field. By drawing Sankey map, combing ICRP’s past publications, and investigating history of radiation protection, this study deeply reviewed the focus of ICRP’s work in different periods, evolution and significance of specific concepts, gradual clarification and deepening of radiation protection concepts, and key directions of ICRP’s future work. By investigating the development and reasons behind the evolution of the classification of radiation situations and the underlying objectives in radiation protection over different periods, this study emphasizes that the shifts in the concept of radiation protection are rooted in the experiences gained through continuous practical applications in real world. Moreover, we provide a detailed discussion on the four major ethical values of radiation protection — beneficence and non-maleficence, prudence, justice, and dignity — highlighting their manifestations and developments during different time periods. This study delves into the relationships between the concepts and principles of radiation protection. Meanwhile, by discussing the relationship between updating of radiation protection theory and the understanding of radiation effects, re-examining the Linear No-Threshold hypothesis based on dose, balancing radiation harms with social values, and standardized management of radiation medicine, the inspiration of the continuous development of ICRP radiation protection concept was proposed.
Research on Measurement and Analysis Method of Radiation Test Section Based on Neural Network
Authors: Lee Jinlin , Zhang Yunsheng , Cheng Jie , Fan Guangming , Jin Shuai
Published: November 2024

In the irradiation test equipment, frequent operations such as material replacement, irradiation device surface replacement, and entry and exit of the reactor are carried out. These factors may lead to sensor failures, resulting in inaccurate test results. During the irradiation test process, the fault diagnosis is usually carried out by operators by observing the parameters during operation and combining their own knowledge. The advantage of using neural networks for fault diagnosis lies in their direct and effective statistical analysis and information extraction of massive, multi-source, and high-dimensional data. This method collects different types of data as the basis, uses the implicit effective information to characterize the operating mode and status of sensors, and thus achieves the purpose of detection and diagnosis. In this study, a neural network fault diagnosis model was established to detect and diagnose the faults of sensors. The fault diagnostic model is trained based on experimental data, and the model is synchronously trained online during the irradiation test process to continuously improve diagnostic efficiency. The fault diagnosis model adopts an unsupervised autoencoder, which takes experimental data as the target for learning. The model trained using obtained from irradiation experiments performs well in simulation tests, and can rapidly and accurately determine the data and process related experimental data when sensor faults occur, providing the location and type of sensor faults, and intuitively and quickly prompt operators with effective information.
Thermal-Hydraulic Analysis of the Upper Plenum in a Typical LFR During a Reactivity Insertion Accident
Authors: Qiu Hanrui , Zhang Ruibo , Zhang Jing , Wang Mingjun , Tian Wenxi , Su G. H.
Published: November 2024

MYRRHA is the Gen IV liquid metal cooled fast reactor, which is a pool-type reactor system developed and designed by the Belgian Nuclear Research Center SCK·CEN. Regarding the safety assessment of the installation, the major challenge is to analyze the complex coolant flow field characteristics in the large open regions of the reactor vessel. In this paper, a coupled sub-channel/CFD multi-scale platform is developed through implementing the sub-channel code SACOS-PB to the CFD code FLUENT based on the domain decomposition and explicit numerical scheme. The flow field characteristics in the upper plenum of reactor emergency shutdown after a reactivity insertion accident (RIA) were investigated. Based on the RIA analysis, the thermal stratification effect has been observed in the upper plenum. The analysis results well demonstrated the 3D flow and local phenomenon of the fluid in the upper plenum, which is significant to reactor design and reactor safety analysis of LFRs.
Prediction of Thermal-Hydraulic Parameters for Autonomous Load Following LFR Core Using Surrogate Model
Authors: Zhang Kefan , Duan Wenshun , Zhang Junjia , Chen Hongli
Published: November 2024

The concept of autonomous load following refers to the reactor’s ability to adjust core power through inherent reactivity feedback, without the need for control rod operation. One promising candidate for this concept is the lead-based fast reactor. Ensuring the safety of the reactor core is a primary concern during the load following process. Given the fixed geometry and similar boundary conditions, surrogate modeling is a suitable approach for calculating maximum temperatures under different quasi-static states. It provides near-accurate results in significantly less time compared to CFD analysis. In this paper, a parametric batch study is conducted using CFD to generate the training dataset with different inlet temperature, mass flow rate, and power. Various surrogate modeling methods are employed for training the model, including the Kriging method, the Gaussian Process method, the Radial basis function method, the Response surface method, and the neural network method. The root mean square error and maximum error of surrogate models are evaluated and compared. Among them, the neural network model demonstrates the best accuracy. Its RMSE is 0.84 K for maximum fuel temperature and 0.24 K for maximum cladding temperature, while the maximum error is 5.65 K for maximum fuel temperature and 0.99 K for maximum cladding temperature.
Numerical Simulation of Agglomeration Characteristics of Liquid Lead-Bismuth Eutectic (LBE) Impurity Particles
Authors: Lu Dingsheng , Yang Yupeng , Pang Xiao , Wang Chenglong , Tian Wenxi , Qiu Suizheng , Su G. H.
Published: November 2024

Liquid lead and lead-bismuth eutectic (LBE) are favored as primary side coolants for lead-cooled fast reactors due to their high heat carrying capacity, density, and boiling point. However, these liquids exhibit strong oxidation effects, leading to the formation of solid lead oxide particles. Additionally, they exert significant corrosion effects on structural materials, with LBE, in particular, demonstrating a lower melting point yet stronger corrosion compared to liquid lead. Consequently, a multitude of impurity particles, resulting from corrosion and oxidation, inevitably exists in liquid LBE. In the dynamic environment of liquid LBE, collisions among particles and between particles and walls can induce particle agglomeration due to fluid flow. The particle agglomeration behavior in liquid LBE significantly influences the particle size distribution (PSD), thereby affecting particle size, location, and motion state. This has implications for the interaction between impurity particles and fluid motion as well as wall surfaces. Through the utilization of numerical simulation, the agglomeration characteristics of impurity particles in liquid LBE can be observed in a more straightforward and intuitive manner. This approach enables a more precise and quantitative description of the particle size distribution (PSD) of impurity particles. The numerical results, offering practical insights, contribute valuable information for the formulation of schemes and measures in practical engineering applications. It serves as a reference for controlling and predicting impurities in lead-cooled fast reactors.
Experimental Study on the Vibration Characteristics of Two-Phase Flow Post Dryout in Helical Coiled Tube
Authors: Wang Ningyuan , Chen Deqi , Li Zhenzhong , Liu Hanzhou , Bu Shanshan
Published: November 2024

Boiling two-phase flow after drying is one of the main flow forms in spiral tubes. Exploring the flow-induced vibration response law caused by it is of great significance to the safety analysis and design of steam generators. Leveraging laser Doppler vibration measurement, this study yields crucial experimental data encompassing post-dryout flow rates, wall and fluid temperatures, and vibration displacements. Post-dryout, the root mean square (RMS) of vibration displacement follows a distinctive trend, initially escalating and subsequently diminishing with the upswing in outlet thermal equilibrium steam content. Pre-dryout, the RMS of vibration hovers around 238 μm, spiking dramatically to 1298 μm post-dryout, eventually stabilizing at approximately 200 μm.

Significantly, a robust correlation manifests between the vibration frequency of helical coiled tube and the thermal equilibrium steam content. Specifically, at a steam content of 0.70 (pre-dryout), the vibration frequency registers around 0.2 Hz; at 0.85, it diminishes to 0.1 Hz, and surpassing 1.0, it sharply rises to 7.4 Hz. Methodically analyzing experimental data through statistical, time domain, and spectral approaches elucidates the intricate relationship between vibration response characteristics and two-phase boiling flow within the tube. These insights offer valuable references for refining the thermal design and optimizing vibration suppression in helical coiled tube Once-through steam, enhancing their practical applicability and operational stability.
Three-Dimensional Refined Burnup Calculation of Helical-Cruciform Nuclear Fuel
Authors: Qianni Duan , Wei Li , Junmei Wu
Published: November 2024

The Helical Cruciform Nuclear fuel (HCF) is an Advanced Technology of Fuel (ATF), a low-enriched uranium fuel has been proposed by Lightbridge. The innovation of HCF lies in the use of metallic materials and the geometry is a petal-shaped cross-section and axial twist. HCF proposes higher requirements and challenges for predicting the three-dimensional neutron physics characteristics in different burnup with high fidelity, especially in the three-dimensional geometric model and burnup region division. An introduction of refined three-dimensional burnup characteristic analysis method for HCF is proposed in which the HCF is divided into several hexahedral solid burnup regions and the geometric model established based on DAGMC. OpenMC is used for the coupled neutron transport and burnup calculations. The results show that in the circumferential and axial directions, neutron physics variables are inhomogeneous and asymmetric due to the petal-shaped cross-section and the axial twist. There is a greater effect by twist on the axial power density at convex position compared to concave position. With increasing burnup time, the change in power density at convex position is larger than that at concave position, which increases the inhomogeneity of circumferential power density. The refined three-dimensional burnup analysis method provides a refined spatial distribution of physical parameters under different burnup, enhancing the potential and accuracy of coupled calculations in neutron physics, thermal-hydraulics, mechanics, and other aspects.
Numerical Simulation and Analysis of Condensation-Induced Water Hammer in Heat Exchangers for Residual Heat Removal System in Nuclear Power Plant
Authors: Li Jie , Xiong Feng , Pan Jiaqi , Tu Ruiyang , Yu Pei , Zhao Jiaming , Chen Zhengyu , Zhang Zihao , Guo Wentao , Wang Shengfei
Published: November 2024

Condensation-induced water hammer can occur due to the sudden condensation of steam in pipelines, resulting in pressure surges that can damage the system components. The aim of the study is to investigate the potential occurrence of water hammer phenomena during the condensation process, which can lead to significant damage to the heat exchanger and surrounding components. We employed FLUENT to simulate an abnormal transient in the heat exchangers for waste heat discharge system using a hybrid heat transfer model. The development of this model takes into consideration the constraints of the Lee model and the hybrid heat transfer coefficient (HTC) model. The simulation results show that when the cooling water and hot steam come into direct contact and condensation occurrs, the liquid level in the pipe rises rapidly, leading to an unstable gas-liquid interface. A liquid bridging structure is formed near the pipe wall, which may induce slug flow and lead to gas pocket collapse water hammer, causing significant pressure fluctuations and posing a safety hazard to the stable operation of the system. Key water hammer parameters are also monitored in this study, providing theoretical guidance for engineering design and accident prevention.
Preliminary Design and Transient Simulation of a Nuclear-Solar Hybrid System Using an Open-Air Brayton Cycle
Authors: Deng Jiaolong , Guan Chaoran , Liu Xiaojing , He Hui , Chai Xiang
Published: November 2024

During the energy transition period, global governments are dedicated to develop renewable energy to address environmental challenges. Solar energy has intrinsic problems such as intermittency and volatility. Conversely, nuclear energy boasts stability and low-carbon features. Therefore, this work proposes a new concept of a nuclear-solar hybrid system (NSHS), which shows promise in improving the grid penetration of renewable energy. A 1 MWt Heat Pipe-cooled nuclear reactor (HPR) is coupled with Concentrating Solar Power (CSP) and Thermal Energy Storage (TES) units using an open-air Brayton cycle (OBC). For the micro HPNR, the point kinetic (PK) model considering temperature feedback effects is used to model nuclear power evolution. For the high-temperature heat pipe, a heat transfer limit-based variable resistance model is developed to represent the main features of this high-efficiency thermal device. A secondary solar salt circuit comprises two TES tanks, a solar receiver, and a secondary heat exchanger. Finally, a composite control strategy is proposed to manage the operation of the hybrid energy system, where the nuclear and solar parts are responsible for base-load and peak-load, respectively. The mentioned components and models of the NSHS are integrally constructed using the MATLAB/Simulink environment. A transient simulation of the NSHS is conducted during a typical sunny day. The reactor thermal power varies from 1 MW to 800 kW due to temperature feedback. The volumes of the two TES tanks change due to varying solar radiation and electricity demands. Over the whole year, the total generated power is 2635.6 MWh, resulting in a total energy efficiency of 22.9%.
Multi-Physics Coupling Simulation of a Small Floating Lead-Cooled Fast Reactor Based on OpenMC and GeN-Foam
Authors: Huang Haochen , Xie Fei , Liu Yu , Lu Daogang
Published: November 2024

The small modular floating lead-cooled fast reactor represents a typical fourth-generation reactor, featuring advantages such as compact structure, flexible layout, and inherent safety, making it suitable for providing power to oceanic islands and remote areas, and thus, is of significant importance for global energy security. This paper focuses on a 50MWth fully natural circulation small modular floating lead-cooled fast reactor, and conducts research on the multi-physics coupling of the reactor. Firstly, utilizing the OpenMC program, simplified model of the small lead-cooled fast reactor is established for simulation and calculation of the reactor’s cross-section parameters, which will serve as input parameters for the GeN-Foam program. Subsequently, a full-core three-dimensional multi-physics field coupling model encompassing neutron physics, thermal-hydraulic, and structural mechanics is established using the GeN-Foam program. The model is fed with the cross-sectional parameters calculated by the OpenMC program and other relevant core parameters, and steady-state operational numerical simulation is conducted, yielding positive findings for parameters like core outlet temperature and maximum fuel temperature. Finally, simulations of transient emergency reactor shutdown scenarios are performed, obtaining the changing patterns of operational parameters such as coolant temperature distribution, core power density distribution, and fuel rod deformation with respect to control rod position and power level. This paper plays a significant role in the design and operation of small lead-cooled fast reactors.
Modeling and Control Scheme Design of a Multi-Modular High Temperature Gas Cooled Reactor Cogeneration Unit
Authors: Cheng Zhonghua , Dong Zhe
Published: November 2024

The Multi-Modular High Temperature Gas-cooled Reactor (MHTGR) is equipped with inherent safety features and boasts numerous advantages, notably its capacity to generate high outlet steam temperatures. By leveraging the potential of the high-temperature gas-cooled reactor for cogeneration, it is possible to fully exploit its strengths and enhance its economic feasibility. Compared to MHTGR power generation unit, the multi-modular high-temperature gas-cooled reactor combined heat and power (CHP) system is more intricate, primarily due to the interconnection between heat load and electrical load. Its operating conditions are more diverse, encompassing the necessity to adapt to heat load fluctuations and to track heat load. Therefore, investigating the dynamic characteristics of the multi-modular high-temperature gas-cooled reactor CHP unit is of great significance for the subsequent design of relevant coordinated control schemes. The focus of the present study is a superheated steam generation system utilizing steam conversion technology. This system extracts heat from the high-temperature reactor main steam to raise the temperature of desalinated water, producing high temperature, high-pressure steam for supply to a chemical industrial park.

In this study, dynamic models were developed for each component and integrated for simulation purposes. The open-loop dynamics of the model complied with the analysis, and the model exhibited good simulation accuracy, with most of the process parameters deviating by less than 15% from the set values. A control scheme for superheated steam temperature and pressure has been designed. The closed-loop responses of the system under two operating conditions confirmed the correctness and effectiveness of the designed control scheme.
Numerical Simulation Study on Axial Mixing Characteristics of Fuel Assembly
Authors: Zhang Xiaoyang , Wei Junhan , Zhao Minfu
Published: November 2024

Because the power in the reactor is distributed in cosine, the CHF is usually not located at the exit. In order to accurately predict the CHF point and increase the accuracy of the reactor safety analysis, it is necessary to conduct in-depth research on the axial mixing coefficient distribution of the fuel assembly. This study used the CFD method to investigate the flow mixing of the rod bundle and spacer grid of fuel assembly. Establish a geometric model for the rod bundle and spacer grid, select the SST model for calculation, and obtain the temperature distribution of sub-channels at different axial positions under rated operating conditions. Using the sub-channel analysis program in conjunction with numerical simulation data, determine the mixing coefficients at different axial positions of rod bundle and spacer grid. The variation law of mixing coefficient along axial position is summarized. The results show that for the rod bundle, the mixing coefficient decreases with the increase of axial length, and shows a tendency of approaching 0. The location of the spacer grid has an important effect on the mixing coefficient. The mixing coefficient of the spacer grid has a significant change compared with that of the rod bundle, but the effect decreases with the increase of the distance. In this study, a method was established to calculate the mixing coefficient of different axial positions, and the variation law of the mixing coefficient along the axial position was obtained, which laid a foundation for the development of the mixing model reflecting the axial difference, and helped to improve the accuracy of safety analysis.
Criteria for Selecting Nuclear Reactors for Merchant Shipping
Authors: Emblemsvåg Jan , Ordóñez César Hueso , Tamm Cristian Garrido , Strand Terje , Thoresen Helge , Ortigosa Javier Santos , Pérez Alejandría , Colilles Yolanda Mugica , Izquierdo Alba Esteban
Published: November 2024

The Deep-Sea Fleet consumes about 300 million tonnes of Heavy Fuel Oil (HFO) annually and emits 3% of global greenhouse gas emissions. Currently, green ammonia is seen as the most promising zero-emission substitute. Yet, if green ammonia was to replace HFO, an equivalent to 2.7 times the total EU electricity production in 2022 would be required to produce this fuel. Therefore, the shipping industry investigates whether nuclear power can provide a viable, safe, and cost-effective solution to this challenge. Some navies have successfully used small Light Water Reactors for decades, but studies identify many challenges including costs and public acceptance related to use of such reactors in merchant ships. However, one study demonstrates that for an AfraMax size tanker, the life-cycle cost savings are expected to be 65–70 million US dollars by using a denatured Molten Salt Reactor assuming that the publicly available information is realistic. With over 80 different reactor concepts identified by yearend 2022 primarily from IAEA’s Small Modular Reactor Handbook, there is a need to develop criteria that can help the shipping industry robustly identify suitable nuclear reactors. This paper presents and discusses these criteria to enable the next step in the process, the actual selection of reactor technologies for merchant shipping.
Effect of Air Impurities on the Characteristics of Helium Discharge at High Temperature and High Pressure
Authors: Yang Chuping , Geng Yinan , Wang Jie
Published: November 2024

The high temperature gas-cooled reactor is one of the fourth-generation reactors, which uses helium as the coolant for its low neutron-absorption cross-section and stable chemical properties. In the reactor, the main helium blower, the active magnetic bearings and some other electrical equipment work in high-temperature and high-pressure helium gas. Therefore, it is necessary to study the discharge characteristics of helium under high temperature and high pressure. In this paper, helium models with air impurities were established by COMSOL Multiphysics. To validate the model, the results of these models were compared with those of the pure helium model and the experimental data. A concept named “reaction chain” was introduced via analyzing the reaction rates to elucidate the effects of the air impurities and describe the underlying mechanisms. We observed that the process of helium discharge can be perfectly described by the reaction chains, just like the nuclear reaction. Besides, the results revealed that the Penning ionization of the He* and air impurities can consume He* constantly and directly, which can shorten the reaction chain of the production of dominant ions. Furthermore, the direct ionization of ground state air impurities molecules can produce more electrons to strengthen the effects of electron avalanche, which will further intensify the reactions. As a result, the presence of air impurities can decrease the breakdown voltage remarkably. Finally, due to the differences of energy thresholds of air impurities, the nitrogen, oxygen, and water vapor impurities will have different effects on the breakdown voltage of helium gas.
Joint Simulation of Pressurizer and Chemical and Volume Control System of PWR Based on Modelica
Authors: Gao Zhibo , Xia Xuefeng , Xiang Ziao , Cheng Jie , Wang Jianjun , Wu Di
Published: November 2024

With the increasing emphasis on concepts such as digital reactors, accurately simulating reactors through simulation software is an inevitable goal. Pressurizer and chemical volume control systems are important equipment for maintaining reactor pressure. If some important indicators can be accurately reflected through simulation software, it will be of great significance. This article establishes the pressurizer and the purification and volume control system parts of the chemical volume control system through OpenModelica, and simulates the model under standard conditions, power increase and power decrease conditions. The analysis of the simulation results proves that the established pressurizer and chemical volume control system model has the function of stabilizing pressure and maintaining purification flow.
Hydrodynamic Characteristics of Single Phase Flowing Through Orifice Plate
Authors: Qiao Min , Deng Xiao , Pan Dawei , Huang Weixing
Published: November 2024

The hydrodynamic characteristics of a single fluid flowing through an orifice plate were experimentally studied. Also, it’s flow behaviors were simulated by using computational fluid dynamics. It was found that the pressure drops of single phase flowing through the orifice is mainly caused by the contraction and expansion of the fluid. With increasing flowrate, the contraction effect of the fluid more obvious is and the larger the pressure drop is. Moreover, the press4ure drop increases exponentially with increasing flowrate. The smaller the orifice diameter, the larger the velocity of the fluid is and the more obvious the contraction effect of the fluid after passing through the orifice is, and thus the larger the pressure drop is. The smaller the orifice relative thickness, the greater the velocity of the fluid along the axis after passing through the orifice is and the stronger the contraction effect of the fluid after passing through the orifice is, thus the greater the pressure drop is. The research on this characteristic of single fluid passing through orifice is expected to provide support for the optimal design of sieve plate packed tower and other large heat and mass transfer equipment.
Research on Monte Carlo-Neutron Point Kernel Integration Method Coupling Calculation Method of Neutron Radiation Field
Authors: Zhang Qi , Gong Heli , Liu Liye , Song Yushou , Cao Qinjian , Yong Jinlong , Hu Jiawen
Published: November 2024

Nuclear-powered icebreakers are the future development direction of icebreakers. Due to the fact that icebreakers operate far away from land, the emergency rescue response after a nuclear accident is long and difficult. Therefore, rapid and accurate calculation of the neutron dose in the reactor radiation field is crucial for emergency response to nuclear accidents. This paper coupled the point kernel integration with the Monte Carlo method, and developed a coupled calculation program based on the coupling method. And on the basis that the point kernel program can realize neutron dose calculation, a point kernel program that can be run in parallel was designed to speed up the operation. A single-body source scenario was firstly designed to validate the coupling program. Compared with the results of the Geant4 toolkit, the relative deviations of the coupling program for the calculation of neutron dose are 5%, and the computation time was saved by nearly 90%, which indicates that the coupling calculation program is more efficient in the computation. A computational analysis of the radiation field in the reactor compartment has been carried out on the example of the Russian nuclear-powered icebreaker Arktika. The results show that the coupled computational method takes 10% of the computation time of the Geant4 toolkit under the same computational conditions, and the maximum relative deviation of the computational results does not exceed 20%, which demonstrates the applicability of the coupled calculation method in the complex radiation field in the reactor compartment of the nuclear-powered icebreaker.