bluuebunny/crossref_metadata_2025_split huggingface dataset:
Numerical Investigation on the Enhancement of Steam-Air Condensation Heat Transfer Outside Spiral Pipes
Authors: Jing Ruihan , Zhang Lu , Bian Haozhi , Shen Xinyi , Zhang Xu , Zhou Shuhang , Peng Xiang 
Published: November 2024

Abstract
When loss of coolant accident (LOCA) or main steam line break accident (MSLB) occurs in the nuclear power plant, a large amount of high pressure and temperature steam will be released into the air space, in HPR1000, a Passive Containment Heat Removal System (PCS) is utilized for extracting heat from containment during reactor accident scenarios. Due to the non-condensable gas near the condensation wall will generate additional thermal resistance, it is crucial to enhance external heat transfer in this situation. In this study, the Computational Fluid Dynamics (CFD) method was employed to investigate the heat transfer characteristics of steam-air condensation outside spiral pipes. The effects of different pipe initial spiral radius range from 0.02 m to 0.2 m, pipe pitch of 1.1d and 2.0d, pipe diameter, also pipe bundle on heat transfer characteristics were analyzed. The results indicate that the spiral pipe has an excellent condensation heat transfer enhancement. For a single spiral pipe, a narrow initial spiral radius will deteriorate heat transfer while enhancing heat transfer after increasing the initial spiral radius. When it comes to spiral pipe bundle, increasing the radial pitch enhances air layer shedding, thereby improving condensing heat transfer ability. However, in terms of large pipe pitch, the impact of increasing the radial pitch is minimal.
A Fuzzy Control Method for Adjusting the Peak Value of Power Axial Distribution Based on Distributed Parameter Model
Authors: Dang Airan , Tu Bowen , Luan Xiuchun 
Published: November 2024

Abstract
The neutron diffusion equation is a distributed parametric equation. The Laplace operator of the equation is equivalent to an infinite order matrix, and the analytic form of the steady state power and the space-time distribution of the power can be obtained. A new reactor power model based on neutron diffusion equation is established, which can be used to study not only the time domain variation of the whole power, but also the axial space power distribution.
When the position of the control rod in the core changes, the peak power will appear in the area where no control rod introduces reactivity, and the movement of the control rod always moves with the position of the peak power, which will lead to the uneven power distribution.
In this paper, the original single-group control rod is changed to a two-group control rod, and a fuzzy controller is designed, which can prolong the time for the peak of axial power to reach the peak of steady-state axial power when meets the overall power load following, and reduce the influence of the axial power distribution deviation in the process of control rod movement, and reduce the influence of the power axial offset in the process of control rod movement.
The simulation results show that after the introduction of the designed fuzzy controller, the power load following shows good performance, the peak position of the axial power distribution is closer to the center height, the value is reduced, and the axial offset also reduced.
Development of a High-Temperature Heat Pipe Simulation Code Using Conjugate Heat-Mass Transfer and High-Speed Compressible Algorithm
Authors: Wang Weixiang , Zhang Kefan , Dong Sifan , Pan Rui , Chen Hongli 
Published: November 2024

Abstract
Due to the excellent passive heat transfer characteristics, the Heat Pipe cooled Reactor (HPR) has been widely studied as an important type of micro-reactor. Heat pipe simulation is an unique and important part of the HPRs multi-physics simulation, which can directly determine the temperature of the core boundaries and whether the fission heat can be transported to energy convert unit or dissipated. Heat pipe simulation involves a coupled heat and mass transfer process of multiple physical fields, which posing great challenges. In this work, a two-dimensional high-temperature alkali metal heat pipe simulation code KMC-HPs is developed based on Finite Volume Method (FVM). The heat conduction model is used to characterize the heat pipe wall and porous wick, and a two-dimensional compressible laminar flow model is used to characterize the vapor channel. The laminar flow model is verified with simulation result of commercial code, which indicates the ability of high-speed compressible flow simulation of KMC-HPs. Both steady and transient heat pipe simulations were performed for code validation, and the simulation results were compared with experiment or existing simulation work, which proved the rightness of physical model and numerical algorithm of KMC-HPs, and its ability to simulate heat pipe with high precision.
Control Rod Cooperative Control Strategy Based on Multipoint Reactor Model
Authors: Tu Bowen , Dang Airan , Luan Xiuchun 
Published: November 2024

Abstract
Based on the axial multipoint reactor equation, the effect of reactivity superposition induced by multiple control rods participating in reactor operation is deeply studied, and based on this, an innovative co-coordinated control strategy of primary and secondary control rods is proposed. In this control strategy, the primary control rod is the traditional reactor control rod, which is used for power control; The newly introduced secondary control rod is always in the core, and does not change the overall power value during movement, but is used to flatten the power distribution at the peak power.
The simulation results show that compared with the core power distribution without secondary control rod regulation, the power distribution at the peak of power based on this method is well flattened. At the same time, the axial offset of the core power is reduced compared with the traditional control method, which successfully reduces the negative influence on the power caused by the uneven axial power distribution in the core. In the process of moving the primary and secondary control rods, the secondary control rods can always cover the position of the peak power in the reference model. The simulation results show that there is almost no delay between the secondary control rod and the primary control rod. To sum up the above conclusions, it can be proved that the power regulation method introduced by the secondary control rod can effectively solve the problem of axial power flattening, and the simulation results show that the response speed of the two control rods is basically the same in the control process, and there is no safety problem caused by the reaction delay of the secondary control rod.
Design and Analysis of Integrated Test Facility Based on Small Break Loss of Coolant Accident for AHPR1000
Authors: Yang Changjiang , Ni Si , Zhan Jingxiang , Geng Yiwa , Wang Guangfei , Wu Yuxiang , Yao Di 
Published: November 2024

Abstract
Nuclear energy has the potential to provide a large amount of green and clean energy throughout the world. However, in the process of nuclear energy development, safety has always aroused a lot of attention, especially after the Fukushima incident, safety has become an important condition restricting the development of nuclear power. In the design process of the third-generation nuclear power plant, the incorporation of passive and active safety concepts has further enhanced the defense-in-depth capabilities and demonstrate the effective safety function in both design base accident and design extension condition. To verify the effectiveness of safety system independently designed for next generation of HPR1000, this paper selected important thermo-hydraulic phenomena from phenomenon identification and ranking table in small break LOCA, designed and built an advanced integrated test facility with height ratio of 1:3, volume ratio of 1:90 and power ratio of 1:52 through proportional analysis method of H2TS. It has the test ability to simulate the natural circulation stage, the pressure relief blowout stage, the injection stage of the refueling water storage tank, and the long-term sump recirculation stage. The test facility adopts full pressure design, with consideration for medium-pressure operation capacity, has fully covered the main loop, safety system, auxiliary system and other important systems of the reactor plant prototype. At the same time, it has the function of coupling operation with the containment vessel test device, which can release the mass energy to the vessel and accept the steam reflux condensation. In terms of isothermal, isobaric and equal physical properties simulation, the test parameters have reached the international advanced level, and can carry out multi-condition tests in design basis conditions and design extension conditions, as well as multi-plant tests in active and passive plant. In addition, the test facility has the ability to provide test data under accident conditions, verify safety analysis procedures, and study the impact of key equipment and main operating parameters on the defense-in-depth function under passive and active systems. The test facility could operate stably under high pressure medium pressure and low pressure. Also, the system is sensitive to small break accidents. The distortion evaluation shows that the parameters of the test facility face a good match with the prototype, which can be used to study the key phenomena and safety characteristics of the safety system and confirm the effectiveness of the safety design of AHPR1000.
AHPR1000: Simplified, Intelligent and Environment-Friendly Advanced Nuclear Power Plant
Authors: Wu Yuxiang , Liu Qianwen , Yao Di , Wang Guangfei , Xu Simin , Lu Yang , Wang Chengcheng 
Published: November 2024

Abstract
Nuclear power plays an essential part in the future low-carbon energy mix. However, the current Gen III+ nuclear power plants are generally not cost competitive, and are required to be more environmentally friendly. The emerging technologies such as intelligent technologies are applied in nuclear power plant in a cautious and slow pace. Taking the above into account, AHPR1000 is developed as an upgraded and innovative model on the basis of HPR1000 by the China National Nuclear Corporation, which aims to provide a mainstream clean energy solution around the year of 2030. Inheriting the combined active and passive safety design philosophy of HPR1000, AHPR1000 develops three major passive safety systems to deal with design basic accidents, and a set of active system with minimal power supply and cooling chain to cope with the design extension conditions, which satisfies the overall safety goal with the lowest cost. Other design features such as a single-layer containment providing protection against commercial aircraft crash, a compact nuclear island layout, and the massive modular design are incorporated. Application of digital and intelligent technologies endow AHPR1000 with the characteristics of ubiquitous perception, centralized control, predictive maintenance, intelligent decision-making, and mobile service, reduction of operation and maintenance workload, to achieve the target of “unattended monitoring, less-attended on duty”. By minimizing environmental emissions and minimizing resource consumption, AHPR1000 is environmentally friendly nuclear power plant without any adverse impact on people and the environment throughout the lifetime of construction, operation, maintenance.
An Online State Estimation Method Based on Quantum Genetic Algorithm for Space Nuclear Reactors
Authors: Zhu Enping , Chai Xiang 
Published: November 2024

Abstract
Space nuclear reactors provide continuous, stable, and powerful energy supply for long-term space exploration missions, their operation being independent of sunlight. However, the experimental data of space reactors is very limited, and there is significant uncertainty in numerical models based on traditional numerical algorithms or machine learning. Therefore, it is necessary to develop an online state estimation algorithm to obtain the true operating status of space reactors. This paper introduces an online state estimation method for space nuclear reactors, designed to mitigate the limitations in accuracy and generalizability associated with machine learning-enhanced models. During the operation phase of the reactor, the machine learning accelerated model remains parallel to the real reactor in time. By using sensors and other measuring devices, the reactor transmits measurement data to the model within a certain time interval. After receiving measurement parameters, a quantum genetic algorithm is used for parameter optimization, adjusting critical coefficients in the model to align computational results with actual measurements, thereby more accurately simulating the real operating state. Critical coefficient determination involves a two-step process: First, a preliminary array of critical coefficients is selected through sensitivity analysis of program parameters; Secondly, recursive methods are used to analyze the sensitivity of these critical coefficients to model modifications and determine the final array of critical coefficients. The study focuses on the improved TOPAZ-II model with heat pipes, employing a traditional numerical algorithm-based model as a benchmark. The results show that the proposed method can effectively correct the results of the machine learning-enhanced model, and the corrected model can more accurately reflect the operating state of the reactor.
Soluble Boron Free Small Modular Pressurized Water Reactor Core Design by Using NECP-Bamboo
Authors: Wang Songzhe , Li Yunzhao , Li Yisong , Li Kang 
Published: November 2024

Abstract
Small Modular Pressurized Water Reactor (SM-PWR) refers to small, flexible and modularizable PWR designs. In this study, a SM-PWR without soluble boron is designed and analyzed by using the PWR-core physics analysis code NECP-Bamboo. The reactivity of the core is completely controlled by using burnable poisons and control rods. (1) The rated full thermal power was selected as 200MW, while the diameter and the height of the core were limited as 1.4m and 1.2m, respectively. In this reactor core, hexagonal fuel assemblies were employed, while both the control rods and detectors can be placed in the center of each selected fuel assembly. (2) The fuel rod is wrapped with wire, which not only enhances the heat transfer process, but also serves as structural support. (3) To find the most appropriate fuel arrangement, iterative manual optimization was performed among core arrangements, control schemes, detector arrangements and fuel assembly design to assess acceptable neutronic and thermo-hydraulic core characteristics that can meet the safety margin requirements. (4) The neutronics effects of different burnable and movable neutron absorbers were also analyzed, which leads to the selections of Gd2O3 and B4C as burnable absorbers and AIC as movable control rods, to achieve the cycle length of 1120 effective full power days and to control the reactivity changes during normal operations and load subsequent operations. (5) In addition, the core shutdown allowance exceeds 5000 pcm in the stuck rod criteria. The design of control rod grouping and operation strategy can reduce the core power peak factor fxy to 1.7 and the assembly average burnup 52.5 GWd/tU while pin burnup limited under 70GWd/tU. The Doppler coefficient, moderator temperature coefficient and power reactivity coefficient are all negative during the entire operation process. The radial and axial power distribution results of the deepest burnup fuel rods and fuel rod power envelope show that there is a significant safety margin in terms of fuel temperature limitation. The calculation of fast neutron flux of pellets provides a basis for the subsequent fuel performance analysis. (6) What’s more, the power capacity of the core under Class I conditions were analyzed, including the maximum line power density, Xenon oscillation power distribution and so on, which provides a reference for the nuclear physical design of a new rod-controlled reactor with high U-235 enrichment and no soluble boron.
Automatic Reading Recognition Method for Nuclear Power Digital Detection Instruments Based on Computer Vision and Deep Learning
Authors: Gao Jiarong , Liu Yongkuo , Ai Xin , Shan Longfei 
Published: November 2024

Abstract
Digital detection instruments are crucial monitoring devices in nuclear power plants, and the accurate recognition of instrument readings is of paramount importance for ensuring the safety, reliability, and economic operation of nuclear facilities. To address the challenge of low recognition accuracy caused by complex environmental factors and random noise affecting digital instrument images, this paper proposes a computer vision and deep learning-based method for automatic reading recognition of nuclear array detection instruments. Firstly, the digital detection instrument image sample set undergoes image preprocessing to generate higher-quality images. Subsequently, the limited digital detection instrument images are fed into a Deep Convolutional Generative Adversarial Network (DCGAN) to generate a large-scale training sample set. Building upon this, the YOLOv5 detection algorithm is employed to train the network using the generated sample set. Ultimately, the trained network achieves automatic reading recognition based on input images. The proposed reading recognition method in this paper completes an end-to-end recognition process, simplifying the recognition steps and exhibiting high recognition accuracy and robustness, achieving an accuracy of up to 98.74%.
Study on the Dynamic Characteristics of the Supercritical CO2 Recompression Brayton Cycle for a Lead-Bismuth Generator Set
Authors: Zhang Yifan , Li Hongzhi , Zhou Yujia 
Published: November 2024

Abstract
Lead-cooled fast reactor (LFR) is one type of Gen-IV nuclear reactor with great development potential. The supercritical CO2 (sCO2) Brayton cycle power generation system has the advantages of compact equipment, high efficiency, and great flexibility. Its cycle characteristics are highly consistent with the application requirements of high efficiency, flexibility, and safety of LFR, and it is regarded as the ideal power cycle of LFR. As one of the difficulties of LFR-sCO2 cycle power system, the study on dynamic characteristics and the control strategies of the whole system is relatively scarce. This work takes a MW-scale LFR-sCO2 system as the research object. The dynamic model and the simulation program were established, and the dynamic characteristics of the system were studied. The results show that the increase in rotational speed can quickly raise the mass flow rate and net power, and the surge margin of the compressor also increases significantly, but the net efficiency decreases. The reduction in flow rate or temperature of lead-bismuth eutectic can effectively reduce the heat load of IHX and net power of the system, but the heat power change rate of the system is much faster with the change of mass flow rate. The variation trends of the key parameters with the mass flow rate and temperature of circulating cooling water are basically similar. The split ratio has a significant impact on the compressor surge margin (SMmc and SMrc). With a 5% increase in α, SMmc decreases from 59.89% to 55.87%, while SMrc increases from 46.61% to 56.23%. To avoid the occurrence of compressor surge, it is necessary to pay attention to the change of surge margin caused by the change of α during the variable load operation of the system.
Experimental Study on Natural Settling Characteristics of Aerosols in Containment
Authors: Li Tao , Gu Haifeng , Wang Hui 
Published: November 2024

Abstract
After a serious accident in the PWR nuclear power station, the radioactive aerosol will be released into the containment with the steam from the reactor core. Due to the huge volume of the containment itself, the mixture of steam, air and other gases and fission products may form complex thermal and hydraulic environments such as temperature stratification and aerosol stratification. At present, there is little research on the influence of various factors, such as pressure and vapour volume fraction, on aerosol sedimentation. Therefore, the evolution behavior of aerosol in the large space of containment needs to be further studied. In this study, the deposition behavior of aerosols in large-scale containment simulator (volume 127m3) was studied experimentally. The experiments were carried out under the conditions of different pressure and vapour volume fraction (including pure air). The multi-point position in the containment was measured with the help of particle size spectrometer and other equipment, and the key experimental data such as decay constant and concentration change were obtained.
The Influence of Seasonal Variation on AP1000 Reliability Under LOOP Accident
Authors: Liu Guanyu , Zhang Mingzhu , Wang Bingbing , Yu Yu , Fenglei Niu 
Published: November 2024

Abstract
The benchmark PSA (Probabilistic Safety Assessment) is mainly used to evaluate the average safety level of nuclear power plants, cannot reflect the current risk level of the plant. The risk level of power plants might change under different plant configurations and environmental factors. At present, passive safety technology is being increasingly implemented in advanced reactors. The safety function of passive system is accomplished mainly based on the natural force, which increases the influence of environmental factors on nuclear power plant reliability. Environmental factors vary significantly with the seasons. Therefore, the impact of seasonal changes on the plant reliability is evaluated to consider environmental factors into the analysis of plant reliability. This paper presents a methodological approach to analyze the impact of seasonal variations on the reliability of the AP1000 nuclear power plant under LOOP (Loss of Offsite Power) accident. AP1000 reactor is a typical passive nuclear reactor. The influence of environmental factors on LOOP accident is obvious, and the importance of LOOP accident to is very significant to the reliability of power plant. Initial event and passive system should be considered seriously in the evaluation. The Fault Tree (FT) tool is employed to calculate the frequency of LOOP accident during different seasons. A combination method of Thermal-Hydraulic (T-H) model and probabilistic model is utilized to calculate the failure probability of the Passive Containment Cooling System (PCCS) physical processes. The FT method is used to combine the system functional failure with device fault. Finally, the nuclear power plant reliability under different seasons is evaluated based on the Event Tree (ET) tool.
Experimental Study of the Heat Transfer Performance and the Crust of the Two-Layer Stratified Melting Corium
Authors: Wang Mengyi , Quan Fengyang , Pei Jie , Li Wei , Yuan Yidan 
Published: November 2024

Abstract
During severe accidents, the molten core will relocate into the lower head of RPV. If the lower head of RPV does not receive sufficient cooling, it will be melted by the decay heat and eventually form a melt pool. To avoid radioactive material leakage due to overheating, the in-vessel melt retention (IVR) is applied by flooding the lower head with cooling water. In this study, a two-layer melt pool experimental system is established to simulate the lower head with IVR. The test section geometry is semicircular slice with cooling water channel at the side wall. Power-adjustable electrical heating rods are uniformly arranged in the lower-layer area. The thermal oil and the binary nitrate are simulants of the upper layer and lower layer respectively. The complex heat transfer in the two layers and between the outside cooling water is experimentally studied. It is found that the temperature in the lower-layer area is obviously higher due to electrical heating. As the temperature in the upper-layer area is below the melting temperature, there is crust at the interface. Besides, the heat flux density through the side wall is much higher at the upper layer due to stronger convective heat transfer.