bluuebunny/crossref_metadata_2025 huggingface dataset:
Research on Natural Frequency Characteristics of Spiral Tube Bundles
Authors: Li Xiaoxi , Wen Jiming , Gao Puzhen , Yin Yibo
Published: November 2024

Liquid metal reactors (LMR), with the significant advantages in high safety and good economic benefits, has obvious advantages and broad application prospects in the fourth generation nuclear power systems. Spiral tube steam generator is one of the most important equipment in LMR, which are composed of spiral heat transfer tubes, inner and outer cylinders, feedwater headers, steam headers, and other structures. Due to their compact structure and high heat transfer efficiency, they are conducive to equipment miniaturization and have been widely used in different reactors in various countries. Due to the impact of the shell side fluid in the steam generator on the tube bundle, it is inevitable to cause flow induced vibration (FIV), resulting in thinning or even rupture of the tube wall, thereby shortening the service life or even failure of the steam generator. Therefore, the study of the vibration characteristics of spiral tube bundles is of great significance. The natural frequency of the tube bundle is an important parameter in FIV analysis. The heat transfer tubes of the spiral tube steam generator are made of spiral winding, and the positions between adjacent tubes are complex and variable. The winding direction between adjacent layers of the spiral heat transfer tubes can be in the same or opposite direction, so their spatial structure is complex. This article simulates and calculates the natural frequency of spiral heat transfer tubes in air and lead bismuth, and studies the influence of geometric structural parameters on their natural frequency. Based on simplifying constraints at both ends of the spiral tube bundles to fixed support, the computational fluid dynamics (CFD) simulation was carried out to reveal the effect mechanism of support number, spiral radius and spiral rising angle, on natural frequency of spiral tube bundles in air. The results indicate that an increase in spiral radius leads to a decrease in natural frequency, while an increase in support number leads to a significant increase in natural frequency, and the spiral rise angle has little effect on the natural frequency. Therefore, we can improve the stability of the spiral tube bundles by increasing support number or reducing spiral radius. Through CFD simulation calculation, the influence mechanism of tube bundle structural parameters on the natural frequency of spiral tube bundles in liquid metal environments was studied. The increase of layer pitch ratio and the same layer pitch ratio led to an increase in natural frequency. The research results provide data reference for the prediction of the natural frequency of heat transfer tubes and the study of FIV of tube bundles, which are of great significance for avoiding the failure of heat transfer tubes.
Numerical Simulation of Turbulent Cross Flow Over Helical Tube Bundles With Supporting Structures
Authors: Ma Anxiang , Xie Xiaoyang , Zhao Houjian , Li Xiaowei , Wu Xinxin
Published: November 2024

Helical tube bundles are usually used for the steam generator of high temperature gas-cooled reactor pebble-bed module (HTR-PM). The helical tube bundle is composed of several layers of helically coiled tubes, which are fixed by the supporting structures for preventing flow induced vibration. In the helical tube bundle, there are ideal flow paths separated by different layers of helical tubes. The blocking area ratio (the ratio of tube area to total area) of each flow path is different due to the helical diameter of each layer tubes, which may result in non-uniform flow. The presence of supporting structures will also affect the blocking area ratio, which may further increase the flow non-uniformity. The cross flow over helical tube bundles is complicated and difficult to predict. In the current investigation, RANS with SST k-ω model is used to simulate cross flow over a helical tube bundle with supporting structures. The influence of the supporting structure on the flow uniformity is investigated. The computational domain consists of 5 layers and 5 rows of helically coiled tubes. The influence of the helix angle is neglected. Due to the symmetry of the structure, only 1/3 part in the circumferential direction of the helical tube bundle is simulated. The velocity at the gap region between different layers of tubes is 19.74 m/s, corresponding to a Reynolds number of 28, 745. The simulation results are averaged over a long time and full length in the circumferential direction. Time-averaged distributions of velocity and pressure coefficient at different positions are obtained and compared. Multiple symmetrical secondary flows form in the z-normal plane and the secondary flows near the supporting structures are small. The averaged streamwise velocities in different flow paths are non-uniform. The flow oblique direction is investigated at different positions. In the region near the supporting structures, the flow direction is almost along the streamwise direction.
Fluid-Structure Coupling Analysis of U-Type Tube Bundle in PWR Steam Generator
Authors: Jiang Yingying , Xia Hong , Sun Lanxin , Li Shuang , Zhang Jinming
Published: November 2024

The structural integrity and safety of a steam generator are susceptible to flow-induced vibration. However, due to the intricate internal structure and thermal dynamics of the pressurized water reactor steam generator. Investigating the flow field characteristics and flow-induced vibration of the U-type tube bundle in steam generator holds pragmatic significance. The investigation delves into the flow field characteristics on both the primary and secondary sides of the U-type tubes, as well as the flow-induced vibration response, employing a multiphase flow model and a multiphysics coupling method. The distribution patterns of fluid pressure, temperature, and vapor fraction within the secondary side of both primary and secondary flow fields were investigated. Additionally, the distribution patterns of thermal stress and deformation under steady flow conditions in U-type tubes were determined. The results indicated that the support plates in the 8th and 9th layers have a significant impact on the thermal stress and deformation of U-type tubes. The temperature field and resultant thermal stress induced by fluid heat transfer on both sides of the U-type tube emerge as crucial factors affecting tube deformation.
Implementation of Domain Decomposition Parallel and gCMFD in MOCP: Method of Characteristics for Pebble-Bed
Authors: Hao Chen , Wen Yu Chen , Wang Yi Zhen
Published: November 2024

MOCP code is developed with Three-dimensional Method Of Characteristic (3D-MOC) for pebble-bed High Temperature Gas-cooled Reactor (HTGR) high-fidelity neutronics calculation. To further enhance its efficiency, domain decomposition parallel and gCMFD method are implemented in MOCP as alternatives to its previous calculation scheme, namely track grouping parallel and Macro-Track Transport Acceleration (MTTA) method. This paper provides parallel implementation details in track tracing, MOC sweeping and gCMFD solving. Parallel load balance is ensured by balancing subdomain volumes in each processor. Notably, Constructive Solid Geometry (CSG) based fine-coarse meshes mapping procedures are highlighted in this work. A simplified cylindrical model and HTR-10 whole-core model are used to demonstrate the performance of newly developed MOCP. It shows that MOCP parallel efficiency is around 80% using 160 MPI processors in the simplified cylindrical model. In HTR-10 whole-core model, the implemented domain decomposition parallel and gCMFD in newly developed MOCP could reduce around 70% core-hours compared with that using previous calculation scheme.
Experimental Study on the Sealing Performance of Core Barrel Flange for Swimming Pool-Type Low-Temperature Heating Reactor
Authors: Yu Song , Zhai Weiming , Cheng Daoxi , Zhou Ping , Ma Xiaoyao , Li Ruizhi , Xing Mingdi , Gao Weilong
Published: November 2024

The core barrel of the swimming pool-type low-temperature heating reactor is completely sealed by the structure of the flange itself. The failure of flange sealing will affect the thermal efficiency and safety of the reactor. In order to study the sealing performance of the core barrel flange, an experimental facility for measuring the leakage rate of core barrel flange was established by our team. The studies include measurements on the sealing capacity of different flange structure and factors affecting the leakage rate of flange. It was found that the leakage rate of the convex flange was smaller than that of the concave flange and the flat flange under the same experimental conditions. Thus, the convex flange is an optimal structure for the core barrel flange. For the core barrel flange with the same structure, the leakage rate increases with the flange surface roughness or internal-external pressure difference. To reduce the leakage rate of the flange sealing surface, the flange surface roughness of Ra3.2 could be selected. Furthermore, the leakage rate depends on the flange surface pressure and the fluid temperature. The experiments show that the maximum leakage rate of the core barrel flange is about 0.0047kg/s.
Influence of Nuclear Power Plant Interface Complexity on Operator Performance: A Modeling Study
Authors: Chen Pu , Tong Jiejuan
Published: November 2024

The main control room of the nuclear power plant is a complex human-machine tightly coupled system with frequent human-machine interactions, resulting in a high percentage of accidents/incidents due to operator factors. High temperature gas-cooled reactors adopt a multi-module design, with multiple reactor modules driving a single turbine. As a result, future high temperature gas cooled reactor - pebble-bed module nuclear power plants (HTR-PM) may require a single operator to monitor two or more nuclear steam supply system (NSSS), which may introduce new human factor safety issues. The rationality and ease of use of the human-machine interface (HMI) can reduce the cognitive load and the probability of error of the operator and ensure the safe operation of the nuclear power plant. In this paper, we conducted experiments on the HTR-PM600 simulator to simulate the main control room operator’s patrol task on two types of interfaces. And we simulate the cognitive behavior of operators in the above tasks based on the queuing network-model human processor (QN-MHP) method. The results of the simulations and experiments show that the QN-MHP method can well model the cognitive behavior of the nuclear power plant main control room operator. Moreover, the optimized HMI significantly reduces the operator’s task completion time and cognitive load, which proves that the QN-MHP method can support the optimization of HMI in nuclear power plants.
Design and Evaluation of Mitigation Measures for Tsinghua High Flux Reactor Loss-of-Coolant Accidents
Authors: Wang Zhuang , Xie Heng , Zhu Gan , Yeh Szu-En
Published: November 2024

Tsinghua High Flux Reactor is a planned new high flux research reactor in China. This paper designs a set of mitigation measures for loss-of-coolant accidents and preliminarily proposes the emergency core cooling scheme for Tsinghua High Flux Reactor. To verify the effectiveness of the current scheme, the corresponding Relap5 model is established, based on which break accidents at four different locations are simulated. The simulation results are presented. The core pressure, mass flow, coolant temperature, etc. in the transient process are studied and analyzed. According to the characteristics of the accidents, all break accidents are divided into five stages: non-shutdown stage, high-injection stage, low-injection stage, stable forced circulation stage and natural circulation stage. The non-shutdown stage is the most dangerous, and thus high core pressure and flow must be guaranteed. Sudo CHF correlations are used to predict critical heat flux. Results show that the minimum departure nucleate boiling ratios of all break accidents are larger than the safety limit of 1.5. Therefore, the current scheme can ensure core safety even if a double-ended rupture accident occurs.
A Review of the Performance Analysis of Plate Type Fuel Elements in High-Flux Reactors
Authors: Song Yue , Xie Heng , Lv Meng
Published: November 2024

High flux reactors, due to the characteristic of high neutron flux, are widely used in diverse areas such as research, development, education, and training. The plate type fuel elements need to be operated in harsh environments and consider the influence of high temperature, high pressure, intense radiation, and rapid coolant flow. These interactions among various physical fields have implications for the thermal, mechanical, material and chemical properties of the plate type fuel elements, and ultimately influence the high flux reactor’s safety performances. Therefore, to ensure the operational safety of high flux reactors, it is important to conduct safety performance analysis of plate type fuel elements under the effect of various physical fields during the design process of fuel elements in high flux reactors.

This paper studies the comprehensive analysis of domestic and international research results on plate type fuel elements and tries to conclude the impact of various factors on the operational performance of plate type elements, such as hydraulic effects, corrosion effects, thermal effects and irradiation effects. The analysis result will provide some references to further design and construction of high flux reactor.
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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Experimental Investigation of Heat Transfer to Supercritical Pressure R134a in Artificially Roughened Tubes
Authors: Wiltschko Fabian , Sipova Monika , Vit Jan , Cheng Xu
Published: November 2024

The environment in supercritical water-cooled reactors (SCWR) is highly corrosive. With increasing time of operation, an oxide layer grows on the surfaces of the fuel rod cladding, which increases the roughness of their surface. The aim of this study is the investigation of the effect of the roughness due to corrosion on the heat transfer to a fluid at supercritical pressure conditions. The drastic variations in the thermophysical properties of SCW can lead to a deterioration or an enhancement of the heat transfer. The correct prediction of the heat transfer deterioration, which can cause sharp temperature peaks on the fuel rod cladding is a big challenge. Until the current day, most experimental investigations are conducted, using hydraulically smooth heat transfer surfaces. In the present investigation, the effect of the surface roughness is investigated systematically with experiments, using R134a as a working fluid. In the first step, heat transfer experiments are conducted in a hydraulically smooth directly heated tube, covering a wide range of operating conditions. In the second step, the experiments were repeated, using a tube with a rough inner surface. Based on the results from the two test tubes, the effect of the surface roughness on heat transfer is evaluated.
Improved Athermal Recombination Corrected DPA Model for Displacement Damage Cross Section Calculation
Authors: Yin Wen , Zu Tiejun , Cao Liangzhi
Published: November 2024

Accurate evaluation of material irradiation damage is the basis for safe and economical operation of nuclear facilities. Displacement damage is an important component of material irradiation damage, and the level of displacement damage in materials is usually described by the displacement damage rate. The estimation of displacement damage rate requires the prior knowledge of displacement damage cross section and displacement function model is the key to the displacement damage cross section calculation. Traditional displacement damage function models have significant errors in the calculation of the displacement damage function near the average atomic displacement threshold energy because they use step function approximately for the calculation of the number of displaced atoms. In order to improve the accuracy of displacement damage cross section calculation, an improved ARC-DPA (iARC-DPA) model is proposed, which uses the power function model near the average atomic displacement threshold energy and original ARC-DPA model in the other energy part. The iARC-DPA model has been used in the calculation of displacement damage cross sections for different incident particles, including electron, neutron and charged particles. The displacement damage cross sections of different particles are calculated and compared with the experimental data. Numerical results show that the iARC-DPA model largely improves the accuracy of displacement damage cross section calculation.
Considerations on Current Methodologies for the Assessment of Engineering Correlations for Heat Transfer at Supercritical Pressures
Authors: Kassem Sara , Pucciarelli Andrea , Ambrosini Walter
Published: November 2024

The design of industrial equipment operating at supercritical pressures, especially in innovative nuclear reactors like Supercritical Water Reactors (SCWRs) at 25 MPa, demands accurate modeling of convective heat transfer. The prediction of fuel surface temperature is vital for safety and regulatory approval of new reactor concepts.

Over the years, numerous engineering correlations have been developed based on experimental data for various fluids at supercritical pressures. Some of these correlations claim high accuracy, but their performance often falls short when applied to different datasets, particularly under deteriorated heat transfer (DHT) conditions. DHT is a complex phenomenon sensitive to boundary conditions, challenging traditional formulations like the Dittus-Boelter equation with property ratio corrections. Even more sophisticated formulations have not significantly improved accuracy.

Within the EU ECC-SMART Project, a comprehensive review of these correlations was performed, evaluating their performance using two diverse datasets covering a wide range of boundary conditions. Great care was applied in avoiding common pitfalls in assessing correlation accuracy, such as the inappropriate use of experimental wall temperature data, which can bias the results and mislead about the most critical variable to be predicted by a correlation when used in a system code.

The paper presents sample results from this assessment, highlighting the potential and limitations of existing formulations. It also addresses cases where multiple solutions for wall temperature arise, even with simple formulations. We consider correlations both in their original form and with optimized coefficients, shedding light on possible enhancements. This study offers insights into the current state-of-the-art in heat transfer modeling for supercritical fluids and suggests directions for future research.
Inquiry on the Effect of Heating Structures in the Prediction of Flow Stability at Supercritical Pressures
Authors: Betti Alessio , Pucciarelli Andrea , Ambrosini Walter
Published: November 2024

Flow stability is an important concern in the design of apparatuses containing fluids with heating and cooling sections. Fluid density changes, even when they are minimal as in singlephase thermosyphon loops, are often responsible for the occurrence of flow oscillations, which are mostly unwanted while striving for stable operating regimes. Nuclear reactors may be prone to these phenomena, something that poses problems in view of the stability of operation requested, inter alia, by the General Design Criterion 12 in Appendix A to the USA 10 CFR 50 law. A design conducive to stable operation and an early detection and suppression of oscillations is therefore in order.

In the frame of a third IAEA Coordinated Research Project on Supercritical Water Reactors (SCWRs), aiming to close some of the open issues in this technology, the problem of the effect of heating structures on the prediction of flow stability at supercritical pressure is now targeted for being attentively scrutinised. In front of the scarce experimental data available that sometimes provide counterfeiting information with respect to the one proposed by models, a varied modelling approach must be used, searching for possibly overlooked ingredients in the currently adopted numerical recipes. In the present paper, different modelling techniques have been adopted, ranging from the use of 1D time-domain and linear in-house programmes, to system codes and CFD. With reference to some experimental data and predictions appeared in literature, an attentive discussion is carried out on the possible reasons for discrepancies between observations and predictions, aiming to contribute to clarify this open issue.
Investigation in Design Technology of Load Follow Without Changing Soluble Boron of Hualong One
Authors: Huaiming Cui , Zhonghao Ning , TongXian Liu , ChenLin Wang , Tian Ya Li , Yun Cai
Published: November 2024

At present, Hualong One adapted Mode-G operation and control strategy, which needs to adjust soluble boron concentration periodically in daily load follow and unavoidably make some waste water generation. In order to enhance the advancement and competitiveness of Hualong One, we have proposed the design of Mode-C operation and control strategy to replace Mode-G operation and control strategy in the Hualong One improvement project, which can choose single variable automatic control mode or dual variable automatic control mode according to the demand of load change. One of the most notable advantages of Mode-C is that the core can realize typical daily load follow without changing soluble boron in dual variable automatic control mode. This paper investigates the design technology of load follow without changing soluble boron of Hualong One in Mode-C operation and control strategy. The main design contents include control rods set design, control strategy investigation and theoretical calculation. The theoretical practice shows that the design technology proposed by this paper can effectively help Hualong One realize typical daily load follow without changing core soluble boron, which can do notable contribution to reductions in waste water generation.
Inverse Uncertainty Quantification for Subchannel Code With PSBT Experimental Benchmark
Authors: Luo Hanyu , Liu Xiaojing
Published: November 2024

Inverse uncertainty quantification analysis for subchannel programs has been conducted to decide uncertainty distribution of important model parameters based on the modified Markov Chain Monte Carlo (MCMC) algorithm using the PSBT void fraction experiment. The posterior probability distribution expression for input parameter uncertainty can be derived using Bayesian principles, typically representing a proportional relationship. However, obtaining the direct normalization coefficient proves challenging, prompting the use of the MCMC algorithm to address this issue. Given the substantial external forward program calculations associated with stochastic methods, the Back-propagation Neural Network (BPNN) is employed to construct a surrogate model, serving as a substitute for intricate forward calculations. To significantly enhance the precision of the surrogate model, an adaptive approach based on relative entropy minimization is implemented to densify training sample points. The uncertainty of key input model parameters (slip ratio and turbulent mixing coefficient) was quantified, revealing that the uncertainty of the slip ratio is the primary source of void fraction uncertainty. After obtaining information on input parameter uncertainty, forward uncertainty analysis based on input uncertainty was conducted. The resulting uncertainty band for void fraction was obtained under a tolerance limit of 95%-95%. The results indicate that the uncertainty band effectively envelops the experimental data for void fraction. Using the statistical mean of parameter uncertainty obtained, the baseline values were calibrated. The predicted results of the model correction values were found to be more accurate than those of the baseline predictions.