bluuebunny/crossref_metadata_2025 huggingface dataset:
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

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

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

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

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

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

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

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

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

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.
Preparation of SiC Composite Coating on the Surface of Graphite Spheres in VHTR
Authors: Yang Hui , Zhao Hongsheng , Zhang Kaihong , Cheng Xing , Li Ziqiang , Wang Yue , Liu Bing
Published: November 2024

Due to various characteristics such as high thermal efficiency, inherent safety, passive safety, and hydrogen production for industrial use, the high temperature gas-cooled reactor (HTR) is considered as one of the most promising technologies in the future. Fuel elements are the core part of a reactor. In the existing high-temperature gas cooled reactor HTR, spherical fuel elements are involved, with graphite accounting for more than 90% of the mass and volume of the fuel element. However, graphite is prone to oxidation in oxidizing atmospheres at a higher temperature than 400°C. Regarding the design of future HTR, accident conditions must be taken into consideration. Under the accident conditions of inlet and outlet water, the graphite in the spherical fuel element is easy to be oxidized. This causes damage to the fuel elements and increases the release of fission products, which threatens the safety of HTR.

In this study, two-step pack cementation is performed to prepare SiC composite coatings on the surface of graphite spheres with a diameter of 40 mm, which improves the oxidation resistance of graphite spheres. According to research, the thickness of coating is about 150 μm after two-step pack cementation, with SiC and Si as the main phase of the coating. However, it is necessary to improve the density of SiC composite coatings needs. After 7 h of oxidation in static air at 1500°C, the mass loss reaches 22.22%. To a certain extent, the compactness of the coating can be improved by the slurry impregnation and Si sol gel method.
Research on Optimization Method of Personnel Evacuation Path Based on Evacuation Time and Radiation Risk
Authors: Zhang Jiemin , Yan Jin , Gao Jiaxuan , Cao Qun , Wu Nan , Xue Na
Published: November 2024

In a radiation scenario, the non-emergency personnel in the site are the closest to the accident site, and the group most likely to be exposed to radiation. The evacuation preparation and evacuation plan for personnel in the nuclear power plant should be fully studied to ensure the evacuation efficiency under accident conditions and reduce the risk of irradiation. For personnel at different initial positions, the evacuation time and possible radiation dose during the evacuation process should be comprehensively considered, and the evacuation path should be reasonably planned to reduce the radiation risk of personnel during the evacuation process. This article uses an ant colony algorithm to optimize evacuation time and radiation risk to calculate and analyze the evacuation path of personnel. The results indicate that a comprehensive consideration of evacuation time and radiation risk in evacuation path planning can achieve the goal of balancing evacuation time and radiation dose, and can provide technical support for decision-making in nuclear emergency protection actions.
Transient Identification of HTR-PM Abnormal Feedwater Flow Based on Graph Neural Networks
Authors: Zhang Wenji , Zhang Tianhao , Li Jitao , Li Duo , Guo Chao , Huang Xiaojin
Published: November 2024

The safety requirements for nuclear power plants are very strict, as an accident may lead to radioactive leakage, causing unbearable losses. Therefore, safety related research in nuclear power plants is very important. High Temperature Reactor-Pebblebed Modules (HTR-PM) are a fourth generation advanced reactor with inherent safety features. HTR-PM adopts modular reactors, using multiple reactors to drive one turbine for power generation. To achieve a multi-module design, HTR-PM has a fluid network for steam and feedwater. When the feedwater flow rate of a modular reactor changes, the physical quantities in the fluid network are also affected, resulting in a very complex transient of the HTR-PM at this time.

In recent years, many deep learning methods have been applied to the fault diagnosis of nuclear power plants. These methods cannot consider the physical structural characteristics of nuclear power plants when extracting data features. Therefore, graph neural networks can effectively solve this problem. The graph neural network constructs a topology graph based on the fluid network structure of HTR-PM and combines it with the topology graph when the features propagate with the network layer, fully considering the physical structural characteristics of the nuclear power plant. After experiments, the network structure of LSTM-GCN has achieved excellent results in transient identification of water supply anomalies.
Research on Accident Prediction of Nuclear Power Plants Based on Deep Learning
Authors: Lv Wei , Li Tong , Wang Bo , Tan Sichao , Li Jiangkuan , Tian Ruifeng
Published: November 2024

Based on the bidirectional long short-term memory network (BILSTM) and the long short-term memory network (LSTM), this paper uses the rolling update mechanism (RU) to obtain the time series training set data and test set data of the original data, and predicts the water level and pressure changes of the Pressurizer after the LOCA accident in nuclear power plants, so as to provide a certain reference for the prediction of LOCA accidents in nuclear power plants. Specifically, the following conclusions are obtained: Compared with the LSTM model, the BILSTM model has better prediction results and less prediction loss. Although the average error of the prediction results of the two models is 10−3 orders of magnitude, the average error of BILSTM is reduced by more than 20% compared with LSTM.
Effect of Non-Condensation Gas on Heat Transfer of Direct Contact Condensation
Authors: Li Shu , Huang Wenxing , Zhang Dandi , Tong Lili
Published: November 2024

The small modular pressurized water reactors adopt a safety relief system to provide overpressure protection for the reactor coolant system. During this process, the mixed gas is discharged into the suppression water tank, causing direct contact condensation. The fraction of the non-condensable gas has a significant impact on the heat transfer. A visual experimental device was established to conduct the experiments of direct contact condensation within the mixed gas mass flux in the range of 177∼425 kg/m2s and the non-condensable gas mass fraction in the range of 0∼20% by using type I multi-holes sparger. The jet plume shapes for the pure steam condensation and mixed gas condensation are investigated in the condensation oscillation zone and jet stable condensation zone. The heat transfer area of pure steam and mixed gas condensation is obtained based on the jet plume shape at the interface. An empirical correlation of the heat transfer coefficient of direct contact condensation with a function of a dimensionless steam mass flux and a driving potential for the condensation process and steam mass fraction is developed and verified with different experimental data. The results show that the empirical correlation of the heat transfer coefficient applies to pure steam and mixed gas, and the error of the model is within ±50%.
Verification and Analysis of Thermal-Hydraulics System Program for a Power Plant Model
Authors: Du Lixin , Zhou Peng , Zhang Hao , Yang Yanhua
Published: November 2024

The thermal-hydraulic system analysis program is mainly used for the calculation and analysis of transient and accident processes in nuclear power plants, and it is one of the important tools for the design safety analysis of pressurized water reactor nuclear power plants. This article introduces a new software package that includes a thermal hydraulic system analysis program, Severe accident Calculation and Evaluation software for Nuclear Energy System (SCENES). SCENES is a thermal-hydraulic system analysis program developed using the staggered grid method based on a two-fluid and six-equations model. To verify the reliability of the thermal hydraulic system analysis program in this software package for calculating transient and accident process conditions in a nuclear power plant, this study models the main hydraulic systems and heat transfer components in the primary and secondary circuits of a nuclear power plant, and simulate and analyze its normal operating and accident conditions separately. The results were also compared and analyzed with the RELAP5 calculation results, which were in good agreement. The results indicate that the SCENES system analysis program is reasonable and reliable in calculating the main thermal hydraulic characteristic parameters of nuclear power plants during steady-state operation and certain accident transient processes. Meanwhile, this study lays the foundation for further program development in the future.
Research on Automatic Cable Layout Method for Nuclear Power Plant Based on A* Algorithm
Authors: Du Hailong , Cheng Jie , Wang Jianjun , Gao Puzhen , Deng Xuchen , Tang Kai , Wang Lei , Su Jincheng
Published: November 2024

The design of cable pathways is indispensable components in the layout design of nuclear power plants. In nuclear power plants, cables exhibit characteristics of being numerous in quantity, diverse in types, and having multiple installation standards. At present, the layout design of cable pathways in a typical nuclear power plant has always been carried out relying heavily on engineering experience. Depending solely on manual deployment not only results in low efficiency but may also not yield the most optimal solutions. To mitigate the human factors on the design and expedite the generation of economically superior solutions, we propose an automatic layout method.

In this method, spatial discretization of nuclear power plant civil structures and equipment is achieved using a grid-based approach, and cable path searches are conducted within the grid map using the A* algorithm. Considering the drawbacks of the traditional A* algorithm’s slow pathfinding in large plants and complex environments, we present an improved A* algorithm tailored for cable layout in nuclear power plants. Firstly, according to the characteristic of prioritizing cable installation along the walls, delineate feature lines along the walls throughout the entire plant area. The feature lines intersect with each other, forming a network that connects various rooms. The A* algorithm prioritizes pathfinding along the feature lines during the routing process, significantly reducing the number of traversed nodes. Finally, the cost function of the A* algorithm has been redesigned to better reflect the actual layout costs. In the traditional A* algorithm, the cost function is designed to assess the distance cost from the start node to the goal node. For the cable layout problem, this cost not only includes the actual length of cable routing but also account for other factors such as bend costs, wall penetration costs, etc.

The aforementioned methods have been implemented in the C# programming language and have undergone testing. The results confirm the feasibility of the proposed method.
Typical Fault Diagnosis Model of Nuclear Power Plant Combined With Knowledge Driven and Data Driven
Authors: Wang Xin , Peng Minjun , Wang Hang , Li Zikang
Published: November 2024

In order to prevent potential radioactive leaks and maintain personnel safety, fault diagnosis in nuclear power plants is crucial. Knowledge-driven and data-driven approaches are two predominant directions in nuclear power plant fault diagnosis. Knowledge-driven approach relies on prior professional knowledge, rules and model for state monitoring and diagnosis, offering strong logic and interpretability. However, it often yields poor results in complex systems. On the other hand, data-driven approach utilizes black-box inference to establish the mapping relationship between complex system characteristics and fault modes. However, this approach lacks interpretability in its operation mechanism, and it is deemed unacceptable for high-security industries such as nuclear power. To address these challenges, this paper introduces the concept of graph neural network and proposes an innovative idea of integrating knowledge-driven and data-driven approaches for fault diagnosis in nuclear power systems. The proposed method embeds the fault propagation path into the data-driven inference process by constructing a graph based on feature nodes and fault type nodes. Furthermore, we analyze how different methods of sampling features and aggregating data influence the diagnostic results during the process of inferring fault paths using graphs with corresponding fault labels. We implement and verify our scheme using a comprehensive simulation simulator for nuclear power devices. The results demonstrate that our method effectively carries out fault diagnosis by combining both interpretable graph-based inference with powerful nonlinear fitting capabilities offered by data-driven approaches.
Influence of Thorium Utilization on the Safety of a Small Modular Molten Salt Reactor
Authors: Wu Jianhui , Zou Chunyan , Yu Chenggang , Zhao Hongkai , Cui Yong , Bao Haotian , Cai Xiangzhou , Chen Jingen
Published: November 2024

Molten salt reactors (MSRs) are one type of reactor system that are characterized by their application of liquid fuel and thorium utilization. Up till now, the performance of thorium-uranium fuel cycle has been extensively studied for MSRs, while the influence of thorium utilization on the MSR core safety is rarely investigated. In this study, a systematic research on the impact of thorium utilization on the core safety for a small modular MSR with a thermal power of 500 WM is conducted. The influence of thorium utilization on the decay heat and radioactivity of spent fuel salt is firstly investigated by comparing the performance of cores fueled with low enriched uranium (LEU) and LEU+Th mixed fuel. It reveals that thorium utilization would raise the decay heat and radioactivity of spent fuel salt due to the yield increase of Pa-233, Sr-93, La-142, La-140, La-143 and Cs-141. Then, the influence of thorium utilization on negative temperature feedback and core delayed neutron (DN) fraction is analyzed, and the results demonstrated that thorium addition to the core will weaken the negative temperature coefficient at the beginning of fuel cycle (BOC) and decrease the core DN fraction, but the negative temperature feedback turns to be better than the core fueled with LEU at the end of fuel cycle (EOC). Finally, the influence of thorium utilization on the response of core to six typical transients, including overcooling/overheating of the inlet fuel salt, increase/decrease of the fuel salt flow rate, reactivity insertion, are studied. The results show that thorium utilization would deteriorate the core response to the transient at BOC, while improve the safety of core at EOC. Most importantly, the safety of the core varies less through the fuel cycle than that of the core fueled with LEU, which would benefit for the design of safety control system.
Transport Characteristics of Two-Phase Flow in Rod Bundle Channels With a Spacer Grid
Authors: Ren Jiaxing , Wang Ruohao , Wang Fangdong , Qiao Shouxu , Tan Sichao , Tian Ruifeng , Gao Puzhen
Published: November 2024

The variation of void fraction and interfacial area concentration can affect the heat transfer, mass transfer, and interfacial drags. This study investigates the air-water two-phase flow in 5 × 5 Rod Bundle Channels with a spacer grid. The four-sensor conductivity probe was used to measure the distribution of local two-phase flow parameters (void fraction, interfacial area concentration, bubble Sauter mean diameter, bubble frequency, and bubble velocity) across channels cross-sections at 36.5, 56, 65.2, 77, and 87.6 hydraulic diameters. The experimental data was used to analyze the development characteristics of two-phase flow parameters across the channel cross-section and along the flow direction, the effect of spacer grid and gas-phase superficial velocities, and the one-dimensional transport of the averaged two-phase flow parameters. The results show that the distribution of the local interfacial structure is mainly affected by the combined effects of the radial forces, such as the lift force, the wall lubrication force, and the turbulent dispersion force. The spacer grid mainly affects the distribution characteristics of interfacial parameters through turbulent vortex aggregation and grid shear fracture, with opposite effects; There is a spacer grid influence zone downstream of the spacer grid, and the distribution of interfacial parameters inside it undergoes drastic changes.
Development and Preliminary Validation of the Fine Fuel Cell Subchannel Analysis Code CUNLUN
Authors: Jiang Hongwei , Chen Guangliang , Tian Zhaofei
Published: November 2024

In response to the low resolution caused by the assumption of homogenization in traditional subchannel analysis methods and the high computational resource consumption of computational fluid dynamics (CFD). This article proposes a novel subchannel calculation and analysis method for pressurized water reactor cores that combines the advantages of CFD and subchannel analysis methods, and develops a thermal hydraulic calculation and analysis code CUNLUN for pressurized water reactor cores based on this method using C++language. Preliminary verification was carried out with the numerical simulation results of ANSYS FLUENT, and the calculation results were basically consistent with the experimental data and FLUENT numerical simulation results, indicating that the code can improve the accuracy of predicting local thermal and hydraulic parameters in the subchannel of the pressurized water reactor core, and can predict the flow characteristics of the coolant in the pressurized water reactor core. This code can provide effective design and research methods for thermal hydraulic analysis of pressurized water reactor fuel assemblies.
Prediction of Sensor Data Accuracy in Thermal Experimental Benches Using GRU-GCN Neural Network Model
Authors: Yang Linjun , Li Tong , Liu Yongchao , Wang Bo , Li Jiangkuan , Wen Jiming , Tan Sichao , Tian Ruifeng
Published: November 2024

As a large-scale facility with multi-system coupling, the operation safety of nuclear power plant is of great importance to the environment and human society. Accurate monitoring of the overall operation status of nuclear power plants (NPP) is not only the key to ensure its safe operation, but also an important means to improve energy efficiency and reduce potential risks. In order to solve the problem of simultaneous prediction of multiple operating parameters of nuclear power plant, this study proposes a deep learning model combining Graph Convolution Network (GCN) and Gated Cycle Unit (GRU), which have the advantages of processing complex network data and time series data respectively. In order to verify the validity of the proposed model, a series of experiments are carried out on the thermal test-bed simulating the operation characteristics of nuclear power plants. The experimental results show that compared with the existing methods, GRU-GCN model has significantly improved the prediction accuracy. This result not only shows the potential of the deep learning model combining GCN and GRU in processing complex system data, but also provides a new idea for the safety monitoring and operation management of nuclear power plants in the future.
Research on the Natural Circulation Ability of Lead-Bismuth-Steamwater Direct Contact Multiphase Flow
Authors: Ma Xiaoyao , Cheng Daoxi , Zhai Weiming , Yu Song , Xing Mingdi , Li Ruizhi , Zhou Ping , Gao Weilong
Published: November 2024

Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR) uses a large amount of steam generated by the direct contact heat exchange between liquid metal and water to drive the natural circulation flow of lead bismuth (Pb-Bi) alloy in the primary circuit. To explore the natural circulation flow characteristics of Pb-Bi-water-steam multiphase flow is of great significance for improving the natural circulation ability of Pb-Bi. This paper establishes a one-dimensional theoretical analysis program and numerical simulation calculation program for the PBWFR system, It’s the law of the natural circulation ability of Pb-Bi was obtained with the change of core power, ascent channel height, water injection subcooling degree and so on. At the same time, the transient process of direct contact between Pb-Bi and water at high temperature and high pressure was simulated by numerical simulation program. The key parameters such as temperature, velocity and void fraction of Pb-Bi multiphase flow in the ascending channel were obtained.

The research in this paper lays a theoretical foundation for in-depth exploration of the natural circulation mechanism of Pb-Bi, which is helpful for the optimal design of PBWFR reactors.
Numerical Analysis of Flow Field and Phase Distribution Characteristics of Sub-Cooled Boiling Flow in the PSBT 5×5 Fuel Bundle Channel
Authors: Liao Hengji , Yan Xiao , Gu Haifeng , Qin Shengjie
Published: November 2024

Rod-bundle coolant channel is the key component of the primary loop in pressurized water reactor. Its local thermal-hydraulic behavior is associated with safety and economic of nuclear power plants. With development of computing power and two-phase flow model, Computational Fluid Dynamics has become one of the most feasible approaches to simulate sub-cooled boiling flow phenomenon of the fuel assembly. In this paper, based on Euler-Euler two-fluid model coupled with wall heat partitioning model, flow boiling performance in the PSBT 5 × 5 full-scale fuel bundle with uniformly heated condition have been adopted to validate this CFD model. Predictions of the sub-channel averaged void fraction profiles were in good agreement with the experimental data. Then, void fraction distribution near the fuel rod and transverse velocity field in the PSBT 5 × 5 rod-bundle channels were analyzed under three axial heat flux profiles. With the increase of total heating power, void fraction on the surface of all the fuel rods increase gradually. Especially, a large number of vapor bubbles were observed to occur near the cold rods located at the certain positions. Comparative analysis of the transverse velocity field downstream from the last mixing-vane spacer grid showed that lateral flow downstream the mixing grid was the main factor that influences phase distribution in the fuel bundle channels. The swirl flow occurred in central sub-channels promoted vapor bubbles in the vicinity of the hot rod to accumulate on center of central sub-channels and induced the liquid phase to continuously wet the heated wall, which would enhance the heat transfer behavior. However, due to the limitation of flow shroud and configuration of the mixing vanes, there was no swirl flow in the center of side or corner sub-channels of the PSBT 5 × 5 rod bundle. Besides, the complex cross flow around the fuel rods in the side or corner sub-channels may guide the liquid away from cold rod surface, causing a substantial increase of near-wall void fraction as the heating power increased and ultimately heat transfer deterioration. The research rules in this paper provide an excellent support for a deep understanding of the evolution of phase distribution from the mixing grid in the 5 × 5 rod bundle.
Reliability Quantification of Passive Containment Cooling System Through Response Surface Methodology
Authors: Shikang Chen , Ronghua Chen , Suizheng Qiu
Published: November 2024

For the new generation of nuclear power plants that widely adopt passive system design, the reliability analysis of passive safety systems shows their important impact increasingly. In addition to hardware failure, functional failure of passive systems is a special and important type of failure modes, which is physical process failure caused by the uncertainty of design functions. This study took the probabilistic safety analysis study of the Passive Containment Cooling System as an example, taking loss of coolant accident as passive system reliability analysis scenario. The functional criteria was based on the pressure within the containment during the long-term cooling period after the PCCS was put into use. The study determined the key operating parameters and design parameters that affect PCCS based on the above scenario and criteria. The study used rank correlation coefficient to accomplish the sensitivity analysis of key parameters on system functions, found out the key influencing factors, and simplified the model. After reducing the number of parameters involved in the study and focusing on key parameters, the artificial neural network (ANN) method was adopted. Based on the best estimation procedure calculation results of design points, the ANN obtained an accurate limit surface with respect to the functional criterion. By quantifying the uncertainty of key parameters, a large number of samples were obtained to complete the response surface calculation, and the failure probability of the system under corresponding accident conditions was obtained. Compared with traditional probability evaluation methods, this research provided a new method to quantify the safety of passive systems, which has higher calculation efficiency and ensured calculation accuracy.
Research on Operation and Control Characteristics of Large Parallel Pipeline Fluid System
Authors: Yuchen Niu , Peng Minjun , Zhang Bowen , Yuan Zhe
Published: November 2024

Large parallel fluid pipeline systems, which exist widely in nuclear power, refrigeration, oil exploitation, and transport as well as other industry, have complex physical coupling relationships among various branches and the main pipeline. This leads to long adjustment cycles of flow and pressure during actual process of conditions changing. Therefore, this study constructed a Modelica language model for an experimental platform of a secondary circuit lubrication system of marine nuclear power plant to investigate rapid adjustment methods for the flow and pressure, thereby enhancing the overall safety and stability. The results indicate that by installing FID controllers on the turbine oil pump and various branch throttling valves, synchronous adjustment of flow and pressure in multiple branches was successfully realized. In comparison to traditional FID controllers, Feedward FID controllers exhibit better resistance to disturbances. FID controllers optimized through the principles of fuzzy control can more rapidly adjust various branch flows to target values, effectively reducing adjustment time. The controller parameter adaptive adjustment scheme can select optimal controller parameters for different condition changing processes.
Design and Performance Verification of Passive Core Cooling System Strainer in Advanced Nuclear Power Plants of China
Authors: Huang Ruolin
Published: November 2024

Once a loss-of-coolant accident (LOCA) occurs in the containment, various materials such as insulation and coating will be damaged by high-energy jet, and the ECCS strainer is used to filter the debris to avoid clogging of downstream equipment. A new advanced nuclear power plant in China equipped with a Passive Core Cooling system (PSI), which relies on the natural circulation driven by the water head difference between sump and core to achieve long-term core cooling after a LOCA. Some pipe break may locate under the sump flood-up elevation, resulting in most of the debris directly entering the core through the break instead of being filtered through the strainer. So a series of debris sources prevention criteria is specified to strictly minimized debris load to avoid clogging the fuel assembly flow path. An ECCS strainer with W-shaped bending filter screen is designed and a strainer module is used as the test prototype to evaluate head loss. According to the test results, 120m2 is selected as the suitable screen area; the chemical precipitate would cause a sharp head loss increase and thus the use of exposed aluminum material should be strictly restricted.
Lattice Physics Study of Various Fuel Assembly Design and Optimization Based on Neutronic Analysis for PT-SCWR
Authors: Shivam Shuvendu , Sekhar Bhogilla Satya , Dutta Goutam
Published: November 2024

The pressure-tube type supercritical water-cooled reactor (PT-SCWR) is a Canadian Generation IV concept advanced reactor designed for higher system pressure and enhanced thermal efficiency. Operating at 25 MPa pressure and temperatures ranging from 350 °C to 625 °C, it achieves over 45% power conversion efficiency. The large variation of coolant density in the core by more than a factor of seven is the driving force for this in-depth core neutronic analysis. Improving lattice parameters in order to achieve the right balance between coolant void reactivity (CVR) and fuel utilization is a major difficulty in the reactor’s physics design. The PT-SCWR features detachable fuel components in pressure tubes immersed in a heavy water moderator. This research focuses on lattice physics and core neutronic analysis, utilizing the DRAGON neutron transport code. Various fuel assembly configurations, including several radial enrichment profiles for different fuels and optimized central rod constituents, were examined for improved safety and efficiency. Innovations, like a two-ring fuel assembly with a central flow tube, enhance fuel exit burnup and reduce coolant void reactivity. The study demonstrates the safety improvements through optimized bundles, meeting criteria for fissile fuel usage and negative coolant reactivity coefficients.
Research on the ODR-VS Method Applicable to Various Geometric Models
Authors: Feng Zhiyuan , Liang Jingang , Guo Wenli , Wang Kan
Published: November 2024

To improve fuel performance and overcome the lack of thermal conductivity and ductility of ceramic nuclear fuel, metal-based fuel elements are used in some high-flux research and test reactors, high-power power reactors and long-life power reactors. However, the stochastic dispersion of fuel particles presents a challenge for accurate modeling and neutron transport analysis. Two commonly employed explicit modeling approaches for stochastic media are the Random Sequential Addition (RSA) method and the Distinct Element Method (DEM). However, the RSA method is constrained by a maximum packing fraction (38.4%). While the DEM method overcomes the limitation of the packing fraction, its inherent complexity, substantial computational requirements, and low packing efficiency remain significant challenges. Moreover, different shapes of fuel elements impose higher demands on geometric universality. Building upon the proposed ODR-VS (Optimized Dropping and Rolling and Virtual Surface) method, we further extend its geometric applicability to include shapes such as cubes, sector cylinders, annular cylinders, and spheres, addressing the geometric modeling needs of a wide range of fuel elements. By employing the enhanced ODR-VS method, we model fuel elements of various shapes and conduct an analysis of the packing efficiency for each type.
Design of a He-Xe Gas-Cooled Nuclear Microreactor System Based on Detailed Models of the Cycle Components
Authors: Guan Chaoran , Chai Xiang , Zhang Tengfei , Liu Xiaojing
Published: November 2024

Nuclear microreactor has disruptive potential as an alternative to carbon-intensive energy technologies based on its mobility, resilience, independence from the grid, long refueling intervals, and low-carbon emissions. A He-Xe gas-cooled nuclear microreactor with the dry-cooled Closed Brayton cycle is designed in this study based on detailed models of the cycle components. Design methods of different components are included. A one-dimensional single-channel model is developed for the reactor core. A one-dimensional discretization is also used for the recuperator, while the two-dimensional temperature field is adopted for the design of the precooler. The turbomachinery uses the mean-line design method. Three He-Xe mixtures are investigated, namely 4 g/mol (pure helium), 15 g/mol, and 40 g/mol. It is found that the system volume is mainly occupied by the reactor core and precooler, and the volume of the turbo-machines is rather small. The 40 g/mol He-Xe achieves the highest thermal efficiency of 42.09% due to the efficiency improvement of the turbo-machines, and the turbomachinery reduces from 5 stages of pure helium to a single-stage of 40 g/mol He-Xe.