bluuebunny/crossref_metadata_2025_split huggingface dataset:
Preliminary Study on Laser Inertial Confinement Fusion Reactor Energy Transfer
Authors: Li Xinze , Chen Ronghua , Zhang Bingqian , Zhang Kui , Zhang Dalin , Tian Wenxi , Qiu Suizheng , Su Guanghui 
Published: November 2024

Abstract
This study focuses on energy transfer within the implosion target of a Laser Inertial Confinement Fusion Reactor (LICFR). The study involves an analysis of heat transfer and tritium production within the Tritium Breeding Blanket (TBM) and the energy cycle of the secondary loop.
The research is based on an operating condition of 200MW. Super critical carbon dioxide (S-CO2) is utilized to cool the first wall of the TBM. Lead-Lithium (PbLi) is employed as both a coolant and tritium breeder material. With the characteristics of this dual-cooling setup, a multi-stage Brayton cycle is established and analyzed. The calculations couple the Monte Carlo code OpenMC with a self-developed heat transfer model. The study investigates the S-CO2 and PbLi dual-cooling blanket models under the 200MW operating condition. Based on the results of these calculations, optimizations are made to the structure and coolant parameters, and a multi-stage Brayton cycle is designed accordingly.
The research results show that the nuclear-thermal coupling model is capable of providing preliminary calculations. Tritium production and energy extraction from the blanket are influenced by both the dimensions of the reactor cavity and the breeding zone.
Three-Dimensional Temperature Field Monitoring of Heat Pipe Reactor Core
Authors: Wang Shidi , Zheng Youqi , Wang Xiayu , Fan Lunhe , Xiao Bowen 
Published: November 2024

Abstract
The heat pipe reactor is a new type of nuclear reactor based on the principle of heat pipe. It adopts the concept of solid core design, and transfers the fission heat through the matrix to the heat pipe by heat conduction. This type of reactor exhibits good inherent safety and single point failure resistance, and is an ideal candidate for space nuclear power systems and transportable nuclear power supply. This paper studies the core temperature monitoring method of heat pipe reactor based on three-dimensional temperature field reconstruction method. The reconstruction method is based on the value of thermocouple arranged inside the matrix. In each assembly, there are 6 thermocouples. They are used to monitor the temperature of matrix and to predict the temperature of cladding and fuel pins. The temperature field reconstruction method reconstructs the whole temperature field through a small number of discrete measuring points of thermocouple, and then expand the measurement data in large quantities. Using this method, the heat pipe failure and cladding failure were predicted, the specific location of the failed heat pipe is determined by monitoring the relative temperature change of the thermocouples at different positions.
The Variation Law of Aerosol Distribution Coefficient at Different Condensation Rates and its Impact on Heat Transfer of Tube Bundles
Authors: E. Xinnuo , Gu Haifeng , Chen Junyan , Wang Hui , Zhou Yanmin , Sun Qingyang , Yu Jianqun , Wang Xiao 
Published: November 2024

Abstract
When the PCCS (passive containment cooling system) operation, the aerosols in the containment along with the steam condensation migration to the condensate, will remain in the heat exchanger tube or by the condensate carried to the ground inside the containment or pit. In order to assess the change rule of aerosol distribution coefficient and the degree of influence on the PCCS heat transfer, this paper carries out an experimental study on the phenomenon of removing aerosols from the PCCS, and the experimental results show that the higher the rate of vapor condensation, the larger the aerosol distribution coefficient; lateral comparison of the aerosol distribution coefficient, in the experimental process of the aerosol distribution coefficient per unit of time is more stable, which indicates that the behavior of aerosols in the liquid phase and the wall of the tube between the liquid phase and the behavior is a stable process; the heat exchanger tube cooling tube is a stable process. A kind of stable process; heat exchanger tube condensation heat exchanger quantity change fluctuation is very small, there is no significant trend of change, the effect of aerosol on the condensation heat exchanger characteristics is not significant.
Research on Heat Transfer Calculation of Ellipsoidal Melt Pool
Authors: Liu Peng , Zhou Tao , Zhang Haolei , Liu Xiaofang 
Published: November 2024

Abstract
The in-vessel melt retention (IVR) technology, as one of the typical technologies to cope with severe accidents in third-generation nuclear power, has been widely used in pressurized water reactor (PWR) nuclear power plants. Based on the computational study of melt pool heat transfer in VVER-1000 under severe accidents, a steady-state heat transfer model of melt retention inside the elliptical lower head pressure vessel was established, and the wall heat flux at each place of the ellipsoidal lower head wall, the thickness of the oxide crust formed by the liquid-phase oxidic layer being cooled and solidified, the residual wall thickness of the pressure vessel, and the failure of the pressure vessel were computed and analyzed. The calculation results show that: the heat flux at the bottom of ellipsoidal pressure vessel lower head is the smallest, the heat flow distribution on the wall surface from the ellipsoidal part to the cylindrical part is discontinuous, and obvious heat focusing effect is observed in the metal layer; the thickness of oxidic crust decreases with the increase of axial height of ellipsoidal melt pool, and the higher the volumetric heat release rate of oxidic pool is, the smaller the thickness of oxidic crust is; the wall of the pressure vessel starts to melt at a certain place of the lower head, and the thickness of wall thins with the increase of heat flux. The wall thickness decreases with the increase of heat flux, and the ablation of the pressure vessel wall increases to the maximum value due to the drastic increase of heat flux in the metal layer area, and the thinner the metal layer is, the smaller the wall thickness of the pressure vessel in contact with the metal layer is, and the greater the probability of failure of the pressure vessel is; therefore, the wall surface of the pressure vessel in contact with the metal layer is the most dangerous position in the case of a serious accident.
Investigation About Classification of Stresses due to Radial Deformation Induced by Design Pressure Applied to the Tubes of Square Pitch Multi-Perforated Portion of Thick Plates
Authors: Duranton Pascal , Bernion Marie , Amzil Aboubakr , Jolly Divjot , Benrabia Abdelhak , Da Silva Tony 
Published: November 2024

Abstract
The design of thick Tube Sheets (TS) of a level 1 heat exchangers in the nuclear industry is a precise and complicated process, and one needs to consider the effects of numerous perforations. It must be noted that the residual stresses from the tube welding and expansion processes are not considered in this study.
The present paper focuses on the stresses induced by the radial deformation due to internal design pressure applied on the inner skin of the tubes expanded into the multi-perforated portion of plates. It is assumed that the stresses due to welding, hydraulic expansion and/or roller expansion have no impact. Indeed, pressure in the perforations tends to expand the multi-perforated portion of the plate. This radial deformation is resisted by the unperforated rim, the surrounding components and by the ligaments themselves.
The present paper studies the radial effect of the pressure in the tube expanded in the Tube Sheet in order to suggest a classification of the induced stress as either primary or secondary.
The multi-perforated portion of plate is often represented by a homogenous equivalent material model, developed to replicate the mechanical behavior of the multi-perforated plate while being far less computationally expensive. These methods consider only the primary to secondary pressure differential across the Tube Sheet in the numerical calculation and not the internal pressure in the tubes expanded in the multi-perforated portion of plate. It can be introduced in the stress results by additional post-processing, usually conservatively or excluded if considered as secondary stresses.
During the Tube Sheet service life, primary stresses, together with, to a lesser extent, secondary stresses, should be limited to avoid the risks of excessive deformation. To assess the risk of excessive deformation, a limit load analyses is performed.
The Finite Element Method (FEM) used is an explicitly modeled perforated zone of a square pitch Tube Sheet with its surrounding components. The tubes having not been explicitly modeled, two approaches are considered to represent the radial load applied to the holes of the Tube Sheet without considering the stiffness of the expanded tubes.
Weight Window Generation Based on Pre-Calculated Response Matrix
Authors: Hu Yingzhe , Shen Pengfei , Jiang Shihang , Huang Shanfang , Wang Kan , Li Zeguang , Liu Zhaoyuan 
Published: November 2024

Abstract
In deep-penetration calculation, the variance reduction method is very important. However, the generation of variance reduction parameters is usually costly and the result is not always satisfying.
In order to balance the accuracy and cost of generation of variance reduction, a lot of variance reduction methods have been proposed and the mesh-based weight window is widely accepted by academic research and industrial calculation. The generation of mesh-based weight window parameters is always a topic in variance reduction research and a lot of Monte Carlo program has implemented different types of weight window parameter generators.
In this paper we proposed a new mesh-based weight window generator based on response matrix iteration prior to the Monte Carlo simulation. And we can show that this method can provide a quick estimation to the flux level of different mesh cells in the weight window mesh thus it can give out the estimated mesh-based weight window parameters for global variance reduction. Considering that Monte Carlo programs have used GVR (Global Variance Reduction) method to update weight window parameters during the Monte Carlo simulation, this method can provide a relatively acceptable iteration start point.
Investigation of Compressible Flow in Natural Circulation Loops With a Non-Boussinesq Algorithm
Authors: Zhang Jinsong , Wu Yongyong , Gui Nan , Zhang Zhen , Yang Xingtuan , Tu Jiyuan , Jiang Shengyao 
Published: November 2024

Abstract
Natural circulation loops are adopted in various engineering fields, including thermal systems, nuclear reactors, and renewable energy systems, whose simplicity, reliability, and energy efficiency are crucial. Understanding and optimizing the performance of natural circulation loops are essential for enhancing heat transfer efficiency and overall system reliability.
In present study, an innovative methodology noted as the Decoupled and Stabilized Lattice Boltzmann Method (DSLBM) is used to the research the compressible physical characteristic of natural circulation loop. It enables the concurrent treatment of pressure, density, and temperature, ensuring the precise retrieval of the entire compressible Navier-Stokes (NS) equation with a constitutive relation at the second-order level. By integrating the equation of state, the temperature equation, and the full NS equation, this method effectively addresses singlephase flow perturbations induced by gravity, pressure fluctuations, and temperature-induced density changes, without relying on the Boussinesq assumptions. This holistic approach offers a comprehensive depiction of the physical mechanisms governing natural circulation loops.
On this basis, the paper analyzes the compressible relationships among heating power, cooling power, flow rate, driving head, and Nusselt number under different working conditions in single-phase natural circulation loops. Through velocity distribution, temperature distribution, and other parameters, the flow characteristics and heat transfer mechanisms of single-phase natural circulation loops can be revealed. The results provide a new insight for the design and optimization of nuclear reactors based natural circulation.
Study on Thermal Conductivity Model for Porous PWR CRUD Depositions
Authors: Liu Yan , Liu Xiaojing , He Hui , Zhang Tengfei , Chai Xiang 
Published: November 2024

Abstract
In this study, corrosion product deposition test is carried out to measure effective thermal conductivity of CRUD depositions. Effective thermal conductivity is found to be a function of heat flux with an average value of 1.3866 W/(m × K). A thermal conductivity model for porous CRUD depositions based on statistically self-similar fractal scaling laws for pore size and capillary tortuousness is also developed to improve high-fidelity performance of CRUD internal heat and mass transfer prediction. The proposed model is expressed as an analytical expression of intrinsic thermal conductivities for solid skeleton and coolant phase, porosity and fractal dimensions for pore and tortuosity. Calculated effective thermal conductivity of CRUD depositions is also compared to the experimental data, showing an error maximum of 0.45 W/(m × K). It is also verified by experimental results in WALT Loop Rod 94 with cladding temperature error of less than 3.45 K and cladding superheat relative error of less than 9.25%. Additionally, the accuracy of the proposed model is compared with the existing models, showing better accuracy of cladding temperature prediction results. The study in this paper could capture microstructure characteristics and help understand heat transfer mechanisms of porous CRUD depositions, further promoting high-fidelity numerical simulation for CRUD depositions and related coupled multi-physical phenomena.
Validation of TRACE Simulation of Small-Break LOCA Test (SB-CL-18) in the Large-Scale Test Facility
Authors: Yagihashi Hideki , Goto Kaho , Sekine Masashi , Tsukamoto Naofumi 
Published: November 2024

Abstract
To evaluate the safety of a nuclear power plant, several accident sequences including design basis accidents are calculated using a plant simulation code, and the performance of the safety measures of the plant is evaluated according to whether it satisfies safety criteria. To demonstrate that these calculations using the simulation code are valid, it is important to validate both the simulation code and the input data. We present our approach to PWR accident analysis based on simulations of a small-break LOCA test (SB-CL-18) in the Large-Scale Test Facility (LSTF) using the three-dimensional thermal hydraulics simulation code TRACE.
To show the applicability of TRACE to small-break LOCA, the simulation of the SB-CL-18 was carried out using TRACE and the simulation results were compared with the measurements for several parameters. Furthermore, the loop seal clearing phenomenon that occurs in small-break LOCAs was investigated through a detailed analysis of measured data and TRACE simulation results.
Comparison of Different Methods to Parameterize Errors in Inverse Modeling of Atmospheric Emissions
Authors: Dong Xinwen , Xu Yuhan , Zhuang Shuhan , Hu Hao , Fang Sheng 
Published: November 2024

Abstract
Inverse modeling has been used to complement and validate the inventory-based estimation of atmospheric emission, and provide crucial temporal information on emissions when an unknown source. It requires the environmental observations and the simulations by atmospheric dispersion models (ADMs) as the input, as well as the a priori emission information but is not mandatory. The parameterization of the errors in observation, ADM, and prior considers the corresponding uncertainty, which affects the performance of an inverse model with respect to the continuity and smoothness in the estimation. Three strategies were investigated for this task in this study, assuming the homologous errors, site-dependent errors, and spatiotemporal correlated site-dependent errors. The standard Tikhonov method and the joint estimation method were used to include these strategies and estimate the emission during a field experiment. The results demonstrate that assuming homologous errors brings a simple and effective parameterization, although the temporal oscillatory in estimation. The strategy of site-dependent errors improves the continuity of estimation with the cost of reduced fit to the observations. Further considering the spatiotemporal correlations among sites leads to a smoother temporal variation of emissions, which achieves the best agreement and correlation with observations based on the joint method.
Development Progress of Neutronics and Thermal-Hydraulics Calculation Program NECP-Panda for Pebble-Bed High Temperature Gas-Cooled Reactor
Authors: Wu Yuxuan , Wang Yongping , Qin Shuai , Cao Liangzhi , Wu Hongchun , Luo Yong 
Published: November 2024

Abstract
The Pebble-Bed High Temperature Gas-cooled Reactor (PB-HTGR) is considered one of the most promising next-generation nuclear energy systems. Due to its unique operational characteristics and neutron spectrum, PB-HTGR presents many challenges and issues in technical improvements and design analysis. To address these demands, the Nuclear Engineering and Computational Physics (NECP) Laboratory at Xi'an Jiaotong University is autonomously developing a neutronics and thermal-hydraulic analysis code suitable for PB-HTGR, named NECP-Panda. Currently, the development of the neutronics calculation module has been completed, while the remaining calculation modules are under development and testing. Aiming to validate the neutronics calculation accuracy of NECP-Panda, the HTR-PM was selected for the first criticality calculation and absorber worth calculation. These calculations were compared with Monte Carlo continuous energy solutions and experimental results. The numerical results demonstrate that NECP-Panda is capable of predicting the loading number of mixed pebbles at first criticality with good accuracy, differing from the experimental results by approximately 800. Additionally, the absorber worth calculation results are in good agreement with the Monte Carlo solutions and measured values. These findings confirm that NECP-Panda possesses the capability for neutron physics calculations appropriate to PB-HTGR, with high calculational precision.
Calculation of Radiation Field and Shutdown Dose Rate for Fusion Reactor Based on cosRMC
Authors: Liu Shichang , Wang Zhenyu , Che Rui 
Published: November 2024

Abstract
For the efficiency and memory bottleneck issues of full reactor high resolution calculation for fusion reactors, this paper carries out the research and program development of an efficient variance reduction algorithm, adaptive mesh algorithm, hybrid parallel algorithm and unstructured mesh-based shutdown dose rate calculation algorithm, which can improve the calculation accuracy of operational and shutdown radiation fields for fusion reactors. The new methods developed in this paper improve the computational efficiency of the Monte Carlo particle transport calculation in the deep penetration problem of fusion reactors, effectively overcome the memory consumption problem of high-resolution calculations for the full reactor, improve the accuracy of shutdown dose rate calculations under complex geometry, and provide efficient computational methods and tools for nuclear design and safety analysis of fusion reactors.
Research on Adaptive Updating Method of Nuclear Power Plant Transient Models Based on Concept Drift
Authors: Li Jitao , Wu Zijian , Huang Xiaojin 
Published: November 2024

Abstract
The transient in Nuclear Power Plant (NPP) can be considered as the process of the system transitioning from one condition to another. By monitoring the time series data during the operation of NPP and utilizing data-driven machine learning approach to build classification and prediction models to identify and predict the transient, it can provide early warnings of abnormalities, assist operators in making decisions in advance when accidents are in the development stage, thereby improving the safety of the NPP. Traditional data-driven machine learning method for transient identification in NPP mainly rely on offline static data for training and use online dynamic data for prediction. However, during the operation of NPP, due to the impacts of environment, operation, working conditions, and equipment aging, the time series data will fundamentally change over time, leading to a shift in the joint distribution between the predictive model and the response variable, known as concept drift. This change significantly reduces the predictive capability of the model. To equip machine learning models with the capability for adaptive updating to address this challenge, the Concept Drift Window Retraining (CDWR) method is proposed based on statistical method and sliding window mechanism. This method can also realize incremental learning with small samples, thereby continuously improving the predictive capability online of the machine learning model. The effectiveness of this method has been verified on a modular high-temperature gas-cooled reactor (MHTGR) simulator.
CFD Analysis of a Four-Loop Pressurized Water Reactor With Multiple Operating Conditions
Authors: Tian Yizhi , Chen Guangliang , Xue Ruojun , Sun Yuchen , Li Jinchao , Li Rui 
Published: November 2024

Abstract
Reactor pressure vessel flow field characterization is an important part of the thermo-hydraulic design of nuclear power plants, which is related to the design of the reactor internal structure, the core inlet flow distribution and the operational safety of nuclear power plants; this paper takes the Siemens KONVOI pressurized water reactor as the research object, and carries out the modeling and large basin simulation calculations for the upper plenum, the lower plenum, the outlet section, the inlet section and the core of the RPV, and adopts the porous medium model for fuel assembly level simulation of the core; Siemens KONVOI reactor-type core operation data and ROCOM experimental data are selected for numerical validation of this study. The salt tracer distribution and cross-mixing were simulated in the form of solving additional scalar equations. The results show that the coolant flow pattern is consistent with the experimental results of the ROCOM unit. Comparison with the calculation results of the subchannel procedure on this basis verifies the accuracy of the thermal parameters of the core and ensures the validation of the established CFD calculation model.
In this paper, the reasonableness of the calculation results of the low plenum and the core is verified by comparing with the existing data, and a polyhedral mesh division scheme suitable for RPV calculations is given to analyse the coolant flow cross-mixing characteristics under different operating conditions, and the calculation results show that the low plenum of the RPV can cross-mix the coolant and distribute the flow rate efficiently, and the flow rate of the external core inlet channel is relatively large, and the coolant flow in the vertical projection of the perforated drum (flow rate distribution device) is relatively large. The flow in the core inlet channel at the vertical projection of the perforated drum (flow distribution device) is relatively small, and the flow distribution in the central area of the core inlet channel is relatively uniform; finally, this paper investigates the transient response characteristics of the flow in the pressure vessel under the accidental condition, and the thermal-hydraulic characteristics of the KONVOI pressurised water reactor under the three-loop operation condition are investigated. The flow distribution characteristics at the core inlet and outlet under asymmetric loop flow conditions are summarised, which provides certain reference value for the optimal design and safety analysis of the reactor.