bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.